ML20206D495

Safety Evaluation Supporting Topical Rept BAW-2251, Demonstration of Mgt of Aging Effects for Rv

ML20206D495
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Issue date:	04/26/1999
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PROJECT-683 NUDOCS 9905040134
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FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION CONCERNING

• DEMONSTRATION OF THE MANAGEMENT OF AGING EFFECTS FOR THE REACTOR VESSEL" BABCOCK & WILCOX OWNERS GROUP REPORT NO. BAW-2251 PROJECT NO. 683 1

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.s t- Table of Contents

1.0 INTRODUCTION

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...........................................1 1.1 Babcock & Wilcox Owners Group Topical Report . . . . . . . . . . . . ............. 1 1.2 Conduct of Staff Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

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SUMMARY

OF TOPICAL REPORT . . . . . . . . . . . . . . . . . . . . . . . - . . . . . . . . . . . . . . . . . 2

2.1 Components and Intended Functions . . . . . . . . . . . . . . . . . . . . . . . . . ........2 2.1.1 Intended Functions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.1.2 Components . . . . . . . . . . . . . . . . .. . . . . ...........................2 2.2 Effects of Aging ............... ................................3 2.3 Aging Management Programs . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . 4 l 2.4 Time-Limited Aging Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.0 STAFF EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1

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3.1 Components and lntended Functions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.1.1 Intended Functions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.1.2 Com ponents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.1.2.1 Reactor Vessel Shell and Closure Head . . . . . . . . . . . . . . . . . . . . . . . 6 3.1.2.1.1 Reactor Vessel Shell . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......7 I

3.1.2.1.1.1 Upper Shell Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.1.2.1.1.2 Shell Assembly (Intermediate and Lower Shell Area) . . . . . . . 8 3.1.2.1.1.3 Lower Vessel Head Assembly . . . . . . . . ................. 8 3.1.2.1.2 Clos.ure Head Assembly . . . . . . . . . . . . . . . . . . . ...............9 i

3.1.2.2 Reactor Vessel Nonies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9 3.1.2.2.1 Inlet Nonles . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 j 3.1.2.2.2 Outlet Nonles . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 l

' 3.1.2.2.3 Core Flood Nonies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..9 l 3.1.2.2.4 incore Instrumentation Noules . . . . . . . . . . . . . . . . . . . . . . . . . . 10  ;

3.1.2.2.5 Control Rod Drive Mechanism Nonles . . . . . . ...... .... . 10 l 3.1.2.2.6 Thermocouple Nonle . . . . . . . . . . . . . .................10 l 3.1.2.3 Reacter VesselInterior Attachments .. ... ... ... .... . .... 11 3.1.2.4 Reactor Vessel Pressure-Retaining Bolted Closures . . . . . . . . . . . . 11 4

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. , e 3.1.2.4.1 Closure Stud Assemblies . . . . . . . . . .. . ... .... ... .. 11 l

3.1.2.4.2 CRDM Nozzle Nut Ring Assemblies and Flange Bolting . . ... . . 11 3.1.2.4.3 Thermocouple Nozzle Nut Ring Assemblies and Flange Bolting. . . 12 3.1.2.5 Components Not Subject to Aging Management Review or Not Within Scope of Report .... . . . .... ... . ... . . . . . . . 12 l

i l 3.1.2.5.1 Components Not Subject to Aging Management Review for Renewal. . 12 l 3.1.2.5.1.1 Closure Head 0-Rings and Fasteners . .,. . .. ........... 12 3.1.2.5.1.2 Monitoring Pipes . . . . . . . . . . . . . . . . . . . ... ... . 13 j 3.1.2.5.1.3 Lifting Lugs . . . . . . . . . . . . .. . . ...... .... . . . 13 3.1.2.5.1.4 Flow Stabilizers , . . . . . ...... .... ..... ... .... . 13 3.1.2.5.1.5 Bolted Reactor Vessel Attachments . . . . . . . . . . . . . . . .14 3.1.2.5.1.6 Seal ledge . . . . . . . . . . . . . . . ... ... .... ..... . 15 3.1.2.5.2 Components Not Within Scope of Topical Report. . . ... .... . . 15 l

3.1.2.5.2.1 Lower CRDM Service Support Skirt . ... .... ... . . .15 3.1.2.5.2.2 Reactor Vessel Support Skirt . . . . . . . ............... .15 l 3.2 Effects of Aging . . . . . .......... . ............ ..... . . . . . . . . . . . 15

( 3.3 Aging Management Programs . . . . . . . . . . ... .. ........ ........ . 17 l 3.3.1 Cracking . . . . . . . . . . . . . . . . . . . . . . . . .... .. ....... .. . . 17 3.3.2 Loss of Material Due to Wastage . . . . . . . . . . . . . . . .. ...... .. . 18 l 3.3.3 Loss of Material or Closure Integrity for Bolted Closures . . . . . . . . . . .18 l 3.3.4 Reduction of Fracture Toughness . . . . . . . . . . . . . ............ . 18 i

3.3.4.1 Reactor Vessel Materials Surveillance Program . . . . . . . . . . . . . . . . . . . 19 3.4 Time-Limited Aging Analyses . . . . . . . . . . ................ .......... . . . . 21 3.4.1 Fatigue (including Environmentally Assisted Fatigue). . . . . . . . ............ 22 3.4.2 Reduction of Fracture Toughness . . . . . . . ..... .... . ............. 24

! 3.4.2.1 Charpy Upper-Shelf Energy ... .... ............... ........... 25 3.4.2.2 Pressurized Thermal Shock . . . . . . . . . . . .. ....... ..... . .... 26 l 3.4.3 Intergranular Separations Under Weld Cladding . . . . . . . . . . . . . .. .... 28

4.0 CONCLUSION

S . . . . . . . .......... . ..... ........... .... ... .. 28 i l 1 4.1 Renewal Applicant Action items . . . . . . ..... . ............ .. . ..... 28 l

5.0 REFERENCES

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4 APPENDIX A. List of Correspondence . . . . . . . . . .. . . . ... . . . . . . . . . .. . . . ..... 33 APPENDIX B. Assessment of Topical Report BAW-2275 . . . . . . . . . . . . . , . . . . . . . . . 34 APPENDIX C. Assessment of Topical Report BAW-2274 . . . . - . . . . . . . . . . . . . . . . 37 1

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FINAL SAFETY EVALUATION

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BY THE OFFICE OF NUCLEAR REACTOR REGULATION CONCERNING

• DEMONSTRATION OF THE MANAGEMENT OF AGING EFFECTS l

FOR THE REACTOR VESSEL" BABCOCK & WILCOX OWNERS GROUP REPORT NO. BAW-2251 PROJECT NO. 683 I

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1.0 INTRODUCTION

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Pursuant to Section 50.51 of Title 10 of the Code of Federal Reaulations (10 CFR 50.51),

licenses to operate nuclear power plants are issued by the U.S. Nuclear Regulatory Commission (NRC) for a fixed period of time not to exceed 40 years; however, these licenses may be renewed by the NRC for a fixed period of time including a period not to exceed 20 years beyond expiration of the current operating license. The Commission's regulations in 10 CFR Part 54, ( 60 FR 22461) published on May 8,1995, set forth the requirements for the renewal of operating licenses for commercial nuclear power plants (Reference 1).

Applicants for license renewal are required by the license renewal rule to perform an integrated plant assessment (IPA). The first step of the IPA,10 CFR 54.21(a)(1), requires the applicant to identify and list structures and components that are subject to an aging management review; 10 CFR 54.21(a)(2) requires the applicant to describe and justify the methods used in meeting the requirements of 10 CFR 54.21(a)(1); and 10 CFR 54.21(a)(3) requires that, for each structure and component identified in 10 CFR 54.21(a)(1), the applicant demonstrate that the effects of l aging will be adequately managed so that the intended function (s) will be maintained consistent l with the current licensing basis (CLB) for the period of extended operation. Furthermore, the l applicant must provide an evaluation of time-limited aging analyses (TLAAs) as required by 10 CFR 54.21(c), including a list of TLAAs, as defined in 10 CFR 54.3.

l 1.1 Babcock & Wilcox Owners Group Topical Reoort l

Sy letter dated June 27,1996, the Babcock & Wilcox Owners Group (B&WOG) Generic License Renewal Program (GLRP) submitted topical report BAW-2251, " Demonstration of the Management of Aging Effects for the Reactor Vessel" (Reference 2), for staff review and approval. The purpose of the topical report is to provide a technical evaluation of the effects of aging of the reactor vessel and demonstrate that the aging effects within the scope of the report are adequately managed for the period of extended operation associated with license renewal. j The topical report provides an individual Babcock & Wilcox (B&W) nuclear power plant utility I owner in the GLRP with the technical details necessary for submitting an application for license renewal.

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f l* s 1.2 Conduct of Staff Review The staff reviewed the B&WOG topical report to determine whether the requirements set forth in 10 CFR 54.2Ha)(3) and (c)(1) were met. The staff also obtained the assistance of Brookhaven

National Laboratory to review areas related to neutron fluence. The staff issued requests for i additional information (RAls) after completing the initial review. The B&WOG respondeu to the staffs RAls. The RAls regarding neutron fluence, are addressed in a separate report BAW-2241P, " Fluence and Uncertainty Methodologies"(Reference 3). Requests for additional information, meeting summaries, and other correspondence are listed in Appendix A. The staff review of Reference 3 is complete. The BAW-2241P safety evaluation was issued on February 18,1999.

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SUMMARY

OF TOPICAL REPORT The B&WOG topical report, BAW-2251, contains a technical evaluation of aging effects related to Babcock and Wilcox (B&W) reactor vessel components, and was provided to the staff to demonstrate that B&WOG member plant owners can adequately manage these effects of aging during the period of extended operation. This evaluation applies to the following B&WOG GLRP member plants:

i Arkansas Nuclear One, Unit 1 (ANO-1)

Oconee Nuclear Station, Units 1,2, and 3 (ONS-1,-2,-3)

Three Mile Island, Unit 1 (TMi-1)

The topical report also contains evaluations of time-limited aging analyses (TLAAs), as defined in 10 CFR 54.3, for the reactor vessel. However, the topical report indicates that the TLAA of flaw growth acceptance prescribed in the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code Section XI inservice inspection (ISI) program (Reference 4) is plant-specific, is not within the scope of the repat, and will be resolved on a plant-specific basis.

2.1 Components and Intended Functions 2.1.1 Intendeo Functions The topical report indicates that the reactor vessel component intended functions that are within the scope of license renewal include:

(1) maintaining the integrity of the reactor vessel pressure boundary in accordance with the l current licensing basis (CLB) '

(2) providing structural support for the reactor vessel internals and core 2.1.2 Components l Section 2.0 of the topical report describes the scope of the report. The reactor vessel shell, I

lower vessel head, nozzles (and safe ends, if provided), interior attachments, and all associated pressure-retaining bolting are within the scope of the report. The reactor vessel internals (i.e.,

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the removable structures and assemblies within the reactor vessel that support the core and direct reactor coolant flow) are not within the scope of the report. Reactor coolant system (RCS) piping attached to reactor vessel nozzles or safe ends, including the welded joints, are not within

the scope of the report and are addressed elsewhere (Reference 5).

The B&WOG divides the reactor vessel items within the scope of the report into four major -

groups based upon the applicable ASME Section XI ISI examination categories (Reference 4):

(1) reactor vessel shell and closure head ..

l (2) reactor vessel nozzles, including safe ends as applicable I (3) reactor vessel interior attachments (4) reactor vessel pressure-retaining bolted closures The topical report indicates that the second closure head 0-ring, 0-ring fasteners, monitoring pipes, lifting lugs, flow stabilizers, alignment keys, key blocks, and seal ledge do not perform any i intended functions and, therefore, are not subject to aging management review for license j renewal. The topical report also indicates that the lower control rod drive mechanism (CRDM) i service support skirt and the reactor vessel support skirt are subject to aging management l

review for license renewal but are not within the scope of the report.

2.2 Effects of Aoina l

j. Section 3.0 of the tocical report discusses the aging effects applicable to the reactor vessel groups describer! soove for the period of extended operation for the participating B&W plants.

The topicr? .4 port states that the following effects of aging could result in adverse impact or loss of any of the reactor vesselintended functions :

+ cracking (initiation and growth)

. - Icss of material

. reduction of fracture toughness loss of mechanical closure integrity (for bolted connections)

Table 3-1 of the topical report provides a detailed list of the subassemblies in each of the reactor vessel groups, and identifies the aging effect applicable to each subassembly, as determined by l

the B&WOG's evaluations. These evaluations included a review ofindustry operating experience to identify past incidents of aging effects applicable to the reactor vessel. This review is discussed in Section 3.5 of the topical report.

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.The following is a summary of Table 3-1 of the topical report:

Maior Reactor Vessel Group Acolicable Aaina Effects Reactor vessel shell and Cracking closure head Loss of material j Reduction of fracture toughness i Reactor vessel nozzles Cracking -.

I Loss of material l Reactor vessel interior attachments Cracking Reactor vessel pressure-retaining Cracking bolted closures Loss of material Loss of mechanical closure integrity 2.3 Aaina Manaaement Procrams Section 4.0 of the topical report discusses the B&WOG bases for demonstrating that the applicable aging effects identified in Section 3.0 of the topical report can be managed by existing programs at ANO-1, ONS-1,-2,-3, and TMI-1 during the period of extended operations of those l plants. Table 4-1 in the topical report provides a detailed summary of the existing programs that I manage aging effects that are applicable to each subassembly of the four major reactor vessel groups identified above. These programs are the following:

. ASME B&PV Code,Section XI, inservice Inspection Program Boric Acid Wastage Surveillance Program (Response to NRC Generic Letter 88-05)

B&WOG Reactor Vessel Integrity Program Technical Specification Leakage Limits 2.4 Time-Limited Aoina Analyses Section 4.5 of the topical report identifies the following TLAAs that are applicr je to the reactor vessel, and presents the B&WOG's proposed aging management programs for each TLAA:

Fatigue (including environmentally assisted fatigue)

Pressurized Thermal Shock Charpy Upper-Shelf Energy Intergranular Separations in Low Alloy Steel Heat-Affected Zones Under Austenitic i Stainless Steel Weld Cladding However, the topical report indicates that the TLAA of flaw growth acceptance in accordance with the ASME Section XI ISI program (Reference 4) is plant-specific, is not within the scope of the report, and will be resolved on a plant-specific basis. j I

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3.0 STAFF EVALUATION The staff reviewed the topical report and additional information submitted by the B&WOG to determine if they demonstrated that the effects of aging on the reactor vessel components i covered by the report will be adequately managed so that the components' intended functions I will be maintained consistent with the CLB for the period of extended operation, in accordance with 10 CFR 54.21(a)(3). This is the last step in the IPA described in 10 CFR 54.21(a).

Besides the IPA, Part 54 requires an evaluation of TLAAs in accordance with 10 CFR 54.21(c).

The staff reviewed the topical report and additional information submitted by the B&WOG to determine if the TLAAs covered by the report were evaluated for license renewalin accordance with 10 CFR 54.21(c)(1).

3.1 Comoonents and Intended Functions 3.1.1 Intended Functions The report indicates that the following reactor vessel component intended functions are within the scope of license renewal as described in 10 CFR 54.4:

(1) maintaining the integrity of the reactor vessel pressure boundary in accordance with the CLB (2) providing structural support for the reactor vessel internals and core The staff agrees with this assessment; however, the staff believes that there is an additional intended function - maintaining the capability to shut down the reactor and maintain it in a safe-shutdown condition. As discussed in Section 3.1.2 below, certain reactor vessel components such as CRDM nozzles are needed to perform this additionalintended function.

By letter dated October 30,1998, the B&WOG responded to the staff's concern. The B&WOG indicated that it does not believe that the function in question is a reactor vessel (RV) component intended function, but that it is a system level function. In addition to the reactor vessel components, a number of components and items are required to shut down the reactor and maintain it in a safe shut down condition including for example: CRDM nozzles, CRDM motor tubes, pressurizer, RCS piping, OTSGs, RCPs, RV nozzles, RV shell, RV internals, decay heat drop line, decay heat pumps, and decay heat coolers. The B&WOG reports are prepared at the component level, and B&WOG expects that the individual licensees, when submitting their respective license renewal applications will describe the systems and their intended functions in their application. Because evaluation of the pressure boundary function will be bounding for l scoping, as compared with the safe-shutdown function, this system level function, need not be added as a reactor vessel component intended function. The staff concurs with B&WOG's assessment.

3.1.2 Components The B&WOG divides the reactor vessel components within the scope of the topical report into four major groupings:

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(1) Reactor vessel shell and closure head upper shell assembly

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. Lower vessel head assembly l

. Closure head assembly (2) Reactor vessel nonles

. Inlet nonles _.

. Outlet noules

. Core flood nonles Incore instrumentation nonles Control rod drive mechanism nonles

. Thermocouple nonles (3) Reactor vessel interior attachments  ;

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. Core guide lugs (4) Reactor vessel pressure-retaining bolted closures I

- Closure stud assemblies

. CRDM nonle nut ring assemblies flange bolting

. Thermocouple nonle nut ring assemblies flange botting The following sections discuss the components evaluated in the topical report. The staff notes that the topical report does not constitute a complete listing of the structures and components subject to an aging management review for the B&WOG GLRP member plants as required by 10 CFR 54.21(a)(1), nor does it describe and justify any methodology for the generation of such a list, as required by 10 CFR 54.21(a)(2). Moreover, the staff did not review information relating to any individual plant to determine whether the topical report accurately reflected its respective reactor vessel designs. Therefore, the staff has made no finding on whether the topical report constitutes the complete list of reactor vessel components for which an aging management review must be done or a scoping methodology submitted as a condition of license renewal.

Individual plant applicants will need to identify the structures and components subject to an aging management review and describe and justify a methodology for developing this list as part of their license renewal applications.

3.1.2.1 Reactor Vessel Shell and Closure Head The reactor vessel shell and closure head are the two major assemblies that provide a pressure i boundary for reactor coolant. They are constructed of low alloy steel clad with austenitic stainless steel, except for a circumferential band in the lower shell that is clad with Alloy 82/182 where the 12 core guide lugs are attached. These components are relied on to l

ensure the integrity of the reactor coolant pressure boundary, and therefore, are within the scope of 10 CFR 54.4(a). The closure head also supports and guides the control rods, and is therefore relied upon to shut down the reactor.

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- 3.1.2.1.1 Reactor Vessel Shell The reactor vessel shell is an approximately 14 foot inner diameter, 37 foot high vertical cylindrical shell with a concave lower head. The reactor vessel shell consists of three sub-assemblies:

(1) Upper shell assembly (upper shell flange and upper shell forgings, also called the

" nozzle belt region")

(2) Shell assembly (intermediate and lower shell areas)

(3) Lower vessel head assembly.

3.1.2.1.1.1 Uooer Shell Assembly The upper shell assembly forms the top third of the reactor vessel. It consists of the upper shell flange, which provides the seating surface for the vessel closure head, and a cylindrical section, which contains the inlet, outlet, and core flood nozzles.

Uooer Shell Flance The upper shell flange is a 30 inch high,16 inch thick clad low alloy steel ring forging. The top horizontal flange surface contains a stainless steel clad mating surface with two concentric grooves for the two 0-ring gaskets used to seal the closure head to the vessel. The O-rings, as discussed in Section 3.1.2.5 below, are not subject to an aging management review for license

- renewal. In addition, there are 60 threaded 6.5-inch diameter holes for the closure studs.

At two locations a small leakage path was machined to provice a flow path from the gap between the two concentric O-rings to the outer side of the flange. Small monitoring pipes are welded to the outlet of these drain holes on the outside of the flange. This drain arrangement permits testing and monitoring for leakage past the O-ring seal. The monitoring pipes, as discussed in Section 3.1.2.5.1 below, are not subject to an aging management review for license renewal.

The inner surface of the flange has a 2.5 inch wide shelf from which the reactor vesselintemals are suspended. This shelf supports the weight of the reactor vessel intemals and the core. A 3- 4 inch wide seal ledge ring, which is used to support the seal plate, is welded on the outside of the vessel flange. This seal ledge, as discussed in Section 3.1.2.5.1 below, is not subject to an aging management review for license renewal.

Uooer Shell Foroinas (Nnnia Belt)

The thick upper nozzle belt region of the reactor vessel is fabricated from two cylindrical ring forgings joined together by a circumferential weld seam. The upper forging is 5 feet,5 inches tall and 12 inches thick. The lower forging is the same height but tapers to an 8.5 inch thickness in the area below the nozzle penetrations. At ONS 1 the two forgings are shorter (approximately 4 feet. 2 inches high each), and a third, shorter, 30 inch high bottom section is made from two 7

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The shell forgings are penetrated by two outlet nozzles, four inlet nozzles, and two core flood .

nozzles. These nozzles are attached to the vessel with full-penetration welds. The interior i surface of the upper shell forging and nozzles are all clad with austenitic stainless steel weld deposit.

3.1.2.1.1.2 Shell Assembfv (Intermediate and Lower Shell Area) ,

The shell assembly consists of the intermediate and lower shells, two 6-foot high, approximately 8.5-inch thick cylinders, joined with circumferential welds. At ONS-2 and ONS-3, the cylinders are both single-piece ring forgings. At the other plants, the cylinders are manufactured from two rollec' alates joined with longitudinal (vertical) welds. The intermediate shells at ANO-1 and TMI-1 have repair welds. The intermediatc and lower shell areas at ONS-1, ONS-2, and' ONS-3 have no repair welds.

The interior surface of the shell assembly is c!ad with austenitic stair.iess steel weld deposit, except for a horizontal band underneath the guide lugs, which is clad with Alloy 82/1G2. The 12 Alloy 600 core guide lugs are welded at equal spacings to the Alloy 82/182 cladding along the bottom of the inner surface of the lower shell assembly. These lugs provide a passive restraint to prevent core drop.

3.1.2.1.1.3 Lower Vessel Head Assembly The lower vessel head is of a semi-hemispherical shape ( i.e., its radius of curvature is larger than the vessel radius). The lower vessel head is made from two pieces: (1) the transition forging, a ring forging for the upper portion; and (2) the bottom head, a formed plate for the center concave region. The two sections are joined by a full-penetration circumferential weld seam. The interior surface of the lower vessel head is clad with austenitic stainless steel weld deposit.

Transition Foraina The transition forging, sometime referred to as the " dutchman forging,"is a ring forging that forms both the transition between the cylindrical vessel shell and the concave lower head and the surface connecting the reactor vessel to the reactor vessel support skirt. The support skirt and its joining weld as discussed in Section 3.1.2.5.2 below are not within the scope of the topical report. The transition forging has the shape of an inverted Y in cross-section. The top leg of the transition forging is welded to the bottom of the lower shell assembly. The inner ring curves inward and is welded to the bottom head. The outer ririg is aligned vertically and is welded to the vessel support skirt.

Bottom Head The bottom head is a concave disc that is welded to the inner leg of the transition forging. The bottom head is penetrated by the 52 incore instrumentation nozzles attached from the inside with partial-penetration welds. The original Alloy 600 incore instrumentation nozzles at all plants 8

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p i j within the scope of this submittal were installed before post-weld heat treatment (PWHT). The l repair that replaced the inner portions of the incore monitoring system (IMS) nozzles was performed in the field after the PWHT.

3.1.2.1.2 Closure Head Assem*.alv )

l The closure head assembly consists of an approximately 6.5 inch thick clad low alloy steel l upper dome (similar to the bottom head) and an approximately 24 inch wide, 30 inch high forged i- flange. The closure head flange is machined to accept 60 closure head studs, which are used to fasten the closure head to the reactor vessel. The closure head has sixty-nine 4.5-inch outside diameter (OD) penetrations for the CRDM nozzles (housings). In addition, the ONS-1 and TMi-1 closure heads contain 8 additional 1-inch penetrations that were intended for i

thermocouples. The Alloy 600 CRDM nozzles and thermocouple nozzles (ONS-1 and TMI-1) I were installed after PWHT.

1 The lower horizontal flange surface has two concentric grooves to accommodate the O-rings and their fastening hardware. Three lifting lugs and the lower CRDM service support skirt are welded to the top of the closure head. The lifting lugs are not subject to aging management review for license renewal, and the lower CRDM service support skirt is not within the scope of l

the topical report. These items are discussed in Section 3.1.2.5 below, i 3.1.2.2 Reactor Vessel Nozzles Reactor vessel nozzles include the inlet and outlet nozzles, core flood nozzles. incore instrumentation nozzles, CRDM nozzles, and, at ONS-1 and TMI-1, thermocouple nozzles. All reactor vessel large bore nozzles (i.e., inlet, outlet, and core flood nozzles) are clad with austenitic stainless steel. These components are relied upon to ensure the integrity of the reactor pressure boundary. The CRDM nozzles also support and guide the control rods, and are therefore relied upon to shut down the reactor.

3.1.2.2.1 Inlet Nozzles The four clad low alloy steel inlet nozzles connect the reactor vessel to the upper cold leg RCS piping to admit flow of reactor coolant cooled in the steam generators back to the reactor core via the discharge from the reactor coolant pumps (RCPs). During normal operation these

, nozzles have a continuous flow of RCS fluid at RCP flow rates with nominal cold leg I

temperatures and reactor coolant chemistry.

3.1.2.2.2 Outlet Nozzles  ;

The two clad low alloy steel outlet nozzles connect the reactor vessel to the hot leg RCS piping to allow flow of primary coolant heated by the reactor core to the steam generators. During normal operation these nozzles have a continuous flow of RCS fluid at RCP flow rates with nominal hot leg tempystures and reactor coolant chemistry.

3.1.2.2.3 Core Flood Nozzles I The two clad low alloy steel core flood nozzles and attached stain less steel safe ends connect 9

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I the reactor vessel to tne 12 inch stainless steel combined core flood / decay heat removal (DHR)/ low-pressure injection (LPI) lines to allow decay heat cooling and emergency cooling l water injection. During normal operation these nozzles are filled with RCS fluid. During reactor / system heatup and cooldown, and during normal shutdown operations, low-pressure, low-temperature (DHR cooler outlet temperature) RCS coolant flows through these nozzles, driven by the DHR pumps. Under emergency core cooling system (ECCS) conditions, borated '

coolant would be injected into the reactor vessel through these nozzles.

A stainless steel flow-restricting venturi is welded inside each core flood nozzle; the purpose of 1 the venturi is to limit break flow in the event of a double-ended-guillotine core flood line break.

3.1.2.2.4 incore Instrumentation Nozzles l

The 52 incore instrumentation no7zles that penetrate the bottom head allow insertion of the incore detector assemblies, which measure neutron flux and temperatures in the reactor core.

These nozzles protrude approximately 12 inches up from the inside bottom head of the vessel and end approximately 8 inches down from the outside bottom head. The nozzles were originally installed as approximately 25 inch long pipes with a 1 inch OD fabricated from single i pieces of Alloy 600. The nozzles were attached to the inner surface of the lower head with partial-penetration welds. Because of flow-induced vibration (FIV) problems experienced during the ONS-1 hot functional testing, these nozzles were subsequently modified in the field to strengthen the portion extending into the vessel interior.

l The original nozzles were cut off above the existing weld inside the lower head, and replaced by 1 2 inch OD Alloy 600 nozzles which were attached by full-penetration butt welds. The original  ;

Alloy 600 incore instrumentation nozzles at all plants within the scope of thic submittal were l installed before final PWHT. The repair / modification that replaced the inner portions of these l nozzles at all plants was performed in the field after final PWHT. Alloy 82/182 weld pads were '

added on the outer surface of the lower head at the instrumentation nozzles penetrations at TMI-1 to facilitate extemal repairs, if required.

3.1.2.2.5 Control Rod Drive Mechanism Nozzles l The CRDMs are aligned and supported by 69 nozzles (these nozzles are referred to as

" housings" in vender documentation) in the reactor vessel closure head. With the exception of TMI-1, the CRDM nozzles are a two-piece design with an Alloy 600 nozzle body and a stainless steel flanged CRDM nozzle adapter joined by a full penetration weld. At TMI-1 the CRDM nozzles are a three-piece design with a forged stainless steel adapter piece between the Alloy '

600 nozzle body and the stainless steel flanged CRDM nozzle adapter.

3.1.2.2.6 Thermocouple Nozzle The closure heads at ONS-1 and TMI-1 have eight 1 inch (3/4 inch Schedule 160) thermocouple nozzles, which are installed by intemal partial-penetration welds outside the region of the CRDM nozzles. The nozzles are a two-piece design with a nozzle tube fabricated from Alloy 600 and a mating flange fabricated from Type F-304 stainless steel. Subsequent design eliminated the thermocouples, and the intemal extension portions of the nozzle were removed by cutting them off 6 inches below the lower surface of the closure head. Stainless steel blind flanges were 10 i

,g installed on the external portions using a bolted configuration. At TMI-1 one of the thermocouple nozzles was subsequently converted for use as a vent line, and another was used to accommodate reactor vessel level instrumentation. }

3.1.2.3 ReactorVesselInterior Attachments Reactor vessel interior attachments include core guide lugs and flow stabilizers. The flow stabilizers are discussed in Section 3.1.2.5.1 below, and are not subject to an aging management review for license renewal. Twelve core guide lugs are welded to the Alloy 82/182 cladding at equal azimuthal distances around the bottom inside surface of the lower shell course. The guide lugs are fabricated from Alloy 600 to achieve thermal expansion compatibility with the reactor vessel shell. The guide lugs provide a secondary core support by limiting the downward displacement of the core and the core support structure in the event of failure of a core support component. These components are relied upon to ensure the capability to shut down the reaGor and maintain it in a safe shutdown cordition.

3.1.2.4 Reactor Vessel Pressure-Retainino Bolted Closurgs The bcl ting materials and bolted closures within the scope of the topical report include the closure stud assemblies securing the closure head to the vessel flange, the CRDM nozzle nut ring assemblies used with CRDM flange bolts to secure the CRDMs to the CRDM nozzles, and ,

the thermocouple nozzle nut ring assemblies used with flange bolts to secure the blind flanges l to the thermocouple nozzles. These components are relied upon to ensure the integrity of the reactor coolant pressure boundary.

3.1.2.4.1.QJpsure Stud Assemblies Sixty closure stud assemblies secure the closure head to the vessel flange. Each assembly consists of a threaded stud 6.5 inches in diameter, a castellated nut, two spherical washers, and a bottom insert. Each stud contains a 1 inch diameter pilot hole down the center of the stud.

The pilot hole enables the use of an elongation-measuring device during stud tensioning. The bottom insert is used to close the bottom of the stud and provide a seat for the measuring device. All items, except for the carbon steelinserts and measuring rods, are made of a nickel-chromium-molybdenum steel.

3.1.2.4.2 CRDM Nozzle Nut Rina Assemblies and Flance Boltina I Sixty-nine CRDM nozzle nut ring assemblies are used with the CRDM flange bolts to secure the l CRDMs to the CRDM nozzles Each assembly consists of two nut ring segments fabricated from low-alloy steel, two 3/8-16 UNC-2A x 1.75 inch long bolts fabricated from carbon steel, and a stainless steel locking clip; Each nut ring segment contains four bolt holes that have 1 1/8 -12 UNF-2B threads, and one bolt hole to accommodate the 3/8 inch diameter bolt used to secure the segment to the bottom side of the CRDM nozzle flange. A locking clip keeps the 3/8 inch I

. diameter bolt in place. The 3/8-inch diameter bolts and locking clip are used to locate the nozzle nut ring assemblies on the CRDM nozzle flange. The pressure boundary structural function is provided by the nut ring assembly and flange bolts. The 3/8-inch diameter bolts and locking clips do not perform a reactor vessel intended functiori in accordance with 10 CFR 54.4(a) and 11

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therefore, are not within the scope of license renewal, and are not subject to aging management review.

3.1.2.4.3 Thermocouple Nozzle Nut Rino Assemblies and Flance Bolt 29 Each thermocouple nozzle has a nut ring assembly with 3/4 inch diameter Sange bolt to secure the blind flange to the thermocouple nozzle flange. Each nut ring assembly cc 1sists of a nut ring fabricated from low alloy steel and two 3/8-16 UNC-2A x 1.25 inch long s ms -

fabricated from carbon steel. Each nut ring contains four bolt holes that have 3/4-10 UNC-2B threads, and two holes to accommodate the 3/8 inch diameter screws that secure the nut ring to the bottom side of the thermocouple nozzle flange. The 3/8 inch diameter screws are used to locate the nozzle nut ring assembly on the thermocouple nozzle flange. The pressure boundary structural function is provided by the nut ring assembly and flange bolts. The 3/8 inch diameter screws do not perform a reactor vessel intended function in accordance with 10 CFR 54.4(a)(1) and therefore, are not within the scope of license renewal, and are not subject to aging management review.

3.1.2.5 Comoonents Not Subiect to Aoino Manacement Review or Not Within Scope of Report in Section 2.5 of the topical report, the B&WOG identifies certain reactor vesselitems that do not perform any intended functions and, therefore, are not subject to an aging management review for license renewal, and provides a basis for the exclusion of each item. The B&WOG also identifies certain reactor vessel items subject to an aging management review for license renewal but not within the scope of the topical report.

The topical report indicates that the second closure head O-ring and 0-ring fasteners, monitoring pipes, lifting lugs, flow stabilizers, alignment keys, key blocks, and seal ledge do not perform any intended functions and, therefore, are not subject to an aging management review for license renewal. In addition, the 3/8 inch diameter fasteners used to secure the CRDM and thermocouple nut ring assemblies are not subject to an aging management review for license renewal, as discussed above in Sections 3.1.2.4.2 and 3.1.2.4.3, respectively. The topical report also indicates that the lower CRDM service support skirt and the reactor vessel support skirt are subiect to an aging management review for license renewal but are not within the scope of the report.

3.1.2.5.1 Components Not Subject to Aoino Manacement Review for Renewal 3.1.2.5.1.1 Closure Head O-Rinos and Fasteners Inner and Outer O-Rinos Two 0.455-inch OD cross-section,0.050-inch thick silver plated Alloy 718 O-rings fit in machined grooves in the closure head and upper vessel shell flanges to provide a leakage boundary. They are replaced whenever the closure head is removed, i.e., at refueling. Therefore, in accordance with 10 CFR 54.21(a)(1)(ii), the O-rings are not subject to an aging management review for license renewal. j i

12 l

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O-Rina Fasteners >

For each O-ring, a set of 12 (8 at ANO-1) fasteners is used to hold the O-ring in contact with the closure head flange. The clips are 1/16-inch diameter stainless steel wires formed into welded loops with a tip that inserts into holes in the outer diameters of the O-rings. The clips are held in place by binding head screws bolted into the closure head at small semicircular recesses along the grooves for the inner and outer O-rings. The screws are held in place by backing plates welded to the upper closure head. The backing' plates may have subsequently been removed at some plants. .The fasteners are used to locate the O-ring on the closure head flange. The pressure boundary structural component is provided by the first 0-ring. These fasteners do not perform a reactor vessel intended function in accordance with 10 CFR 54.4(a)(1) and therefore, are not within the scope of license renewal, and are not subject to aging management review.

3.1.2.5.1.2 Monitorina Pioes Two monitoring pipe connections are welded to the outside vertical surfaces of the upper vessel flange at the two 1/2 inch diameter flow holes. These flow paths penetrate the sealing surface of the vessel flange between the inner and outer O-rings. The 1 inch Schedule 160 stainless steel pipes provide capability to detect leakage of reactor coolant in past the inner 0-ring and a path for pressure testing the O-ring seal integrity; it is not a pressure boundary. The monitoring pipes are used for leakage detection. The pressure boundary structural component is provide by the first 0-Ring. These monitoring pipes do not directly support a reactor vessel intended furrtion, so they are not perform a reactor vessel intended function in accordance with 10 CFR 54.4(a)(1) and therefore, are not within the scope of license renewal, and are not subject to aging management review.

3.1.2.5.1.3 Liftina Luas -

Three low-alloy steel lifting lugs are welded to the upper surface of the closure head. The lifting lugs are used to remove the ciesure head during refueling and servicing. The lifting lugs do not perform a reactor vessel intended function in accordance with 10 CFR 54.4(a)(1) and therefore, are not within the scope of license renewal, and are not subject to aging management review.

3.1.2.5.1.4 Flow Stabilizers Twelve stainless steel flow stabilizers (also known as " flow tuming vanes") are welded to the austenitic stainless steel cladding within the lower-head region. These stabilizers were intended ,

to produce mixing of the fluid in the down comer region as it entered the lower-head region. The j flow stabilizers were originally about 10 inches high; however, due to flow-induced vibration l (FIV) experienced with the incore instrumentation nozzles during the ONS-1 hot functional l testing, these stabilizers have been ground down to a 1-inch height to reduce the velocity of the l flow at the incore monitoring system (IMS) nozzles. l The topical report indicates that the flow stabilizers do not directly support a reactor vessel intended function, so they are not subject to an aging management review for license renewal and are not within the scope of the topical report.

l The staff expressed concems regarding excluding the flow stabilizers from aging management l 13 l

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for license renewal. Although the flow stabilizers themselves do not have safety-related functions, they experienced FIV problems during hot functional testing. The staff's first concem was that the failure of the flow stabilizers could result in FIV and prevent satisfactory accomplishment of the intended functions of the reactor vessel components. B&WOG, in response to the staff concem, indicated that the flow stabilizers were originally installed to promote mixir:g of the downcomer fluid in the lower head of the reactor vessel. B&WOG further indicated that the FlV problem resulted in failure of incore monitoring system (IMS) nozzles.

According to the B&WOG, the flow stabilizers were subsequently ground down in height from 10-inches to 1-inch to reduce hydraulic forces on _the IMS nozzles, and the4MS nozzles were also reinforced. The B&WOG response clarifies that the function of the flow stabilizers was not related to FIV problems and because they were ground down, FIV is not a concem. Therefore, the staff concem is resolved.

The staffs second concem relates to potential cracking of the attachment weld of the flow str bilizer to the reactor vessel. Cracking of the attachment weld could cause the reactor vessel shell to crack, thereby affecting its intended functions. Thus, the flow stabilizers may meet 10 CFR 54.4(a)(2) and be subject to an aging management review for license renewal.

The Statement of Considerations for 10 CFR Part 54,48 Fed. Reg. 22461 (1995), under the section entitled, " Systems, Structures and Components Within Scope of License Renewal," and within paragraph (iii) entitled

• Bounding the Scope of Review," indicates:

"An applicant for license renewal should rely on the plant's CLB, actual plant-specific experience, industry-wide operating experience, as appropriate, and existing engineering evaluations to determine those non-safety-related systems, structures, and components that are the initial focus of the license renewal review.

Consideration of hypothetical failures that could result from system interdependencies that are not part of the CLB and that have not been previously experienced is not required."

The B&WOG has verified that failure of the flow stabilizers has not occurred at any B&W operating plant and considers failure of the flow stabilizers to be a hypothetical failure that is not part of the CLB for the B&W plants. Therefore, the B&WOG recommends the flow stabilizers remain within the scope of license renewal but not subject to an aging management review.

The staff agrees with the B&WOG except that if the flow stabilizers were installed using Alloy 600 and/or Alloy 82/182 weld material, the flow stabilizers are within the scope of license renewal and the applicant must include the flow stabilizers in its Alloy 600 aging management program. Alloy 600 and/or Alloy 82/182 weld materials are susceptible to cracking in primary water environments. If the flow stabilizers are not fabricated using these materials, they are not susceptible to cracking and do not require aging management.

3.1,2.5.1.5 Bolted Reactor Vessel Attachments Alionment Keys i

Four stainless steel alignment keys are doweled and bolted to the inner side of the upper vessel  ;

flange to provide a keying surface to orient the reactor vessel intemals and the closure head. I 14

p They are 11-inch high,6-inch wide,1-3/8-inch thick plates with beveled edges. The alignment keys are fastened with hex socket head cap screws. The closure head key blocks and slots in j l- the reactor vessel internals core support shield and in the upper plenum flanges are designed to

) slide over these keys. The alignment keys do not perform a reactor vessel intended function in i accordance with 10 CFR 54.4(a) and therefore, are not within the scope of license renewal, and l . are not subject to aging management review.

Kev Blocks

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Four s' tainless steel key blocks are doweled and bolted to the bottom of the closure head flange with hex socket head cap screws to mate with the vessel alignment keys. . The blocks are 5 inches high and slotted to mate with the keys. The key blocks do not perform a reactor vessel intended function in accordance with 10 CFR 54.4(a)(1) and therefore, are not within the scope of l . license renewal, and are not subject to aging management review.

3.1.2.5.1.6 SealLedae t

The seal ledge is a 3-inch wide carbon steel ring welded to the outside of the reactor vessel l

upper shell flange. The seal plate fits on top of the ledge and between the outside of the upper l reactor vessel shell fl&nge and the walls of containment biological shield. The seal plate seals the

!. refueling canal. The seal ledge does not perform a reactor vessel intended function in l accordance with 10 CFR 54.4(a)(1) and therefore, is not within the scope of license renewal, and are not subject to aging management review.

j 3.1.2.5.2 Comoonents Not Within Scoot J Topical Reoort 3.1.2.5.2.1 Lower CRDM Service Suooort Skirt L The lower CRDM service support skirt consists of 21 segments forming a slotted cylinder, which 1: welded to the upper surface of the reactor vessel closure head. The lower CRDM service

! support skirt was fabricated as an 8 inch high, 3/4 inch thick,138 inch ID carbon steel cylinder l with a 31/4 inch wide flange on top. This cylinder was welded to the top of the closure head and then 21 vertical sections or slots were cut and ground out, leaving the 21 segment support skirt.

i The flange for each lower skirt segment is penetrated by four vertical bolt holes and two horizontal holes for location dowels. The flange provides the seating surface to which the upper CRDM service support assembly is bolted. The lower CRDM service support structure, including the weld that connects the lower CRDM sarvice support skirt to the reactor vessel closure head, is subject to an aging management revc w for license renewal. The B&WOG has decided to j exclude it from the scope of the topical report. Thus, a renewal applicant needs to address it in j- the license renewal application.

l 3.1.2.5.2.2 Reactor Vessel Sucoort Skirt The reactor vessel support skirt is a 2 inch thick,60 inch high,175.5 inch ID carbon steel ring. It supports the reactor vessel. The vessel support skirt is welded to the bottom of the outer ring of the lower-vessel transition forging. The reactor vessel support skirt, including the weld that connects the reactor vessel support skirt to the transition forging, is subject to an aging management review for license renewal. The B&WOG has decided to exclude it from the scope l 15 I

...- o. o of the topical report. Thus, a renewal applicant needs to address it in the license renewal application.

- 3.2 Effects of Aoina

' As. discussed in Section 2.2 above, the effects of aging evaluated in BAW-2251 included reduction of fracture toughness, cracking, loss of material, and Inss of mechanical integrity (for bolted connections). The B&WOG reviewed these aging effects for their applicability to the RV components within the scope of the report. The B&WOG reviewed RV service history of cracking of weld locations (due to mechanical failure, fatigue and other causes), loss of material (extemal wall thinning), and loss of closure integrity (wear and erosion). The B&WOG findings about these effects were mcorporated into the aging management program.

The staff agrees with the B&WOG identification of applicable RV component aging effects that .

are subject to aging management as a condition of license renewal. The staff had identified two concerns in this area that are resolved as follows:

(1) . Wear of core guide lugs The staff indicated that it considered the loss of materials due to mechanical wear of the core ,

guide lugs a potential applicable aging effect that should be managed for license renewal. This potential aging effect is discussed in Section 3.1 of the working draft standard review plan for license renewal. The B&WOG response indicates that indentations on the sides of selected guide lugs have been observed at one B&W operating plant. The B&WOG further agreed that loss of material at the locations where the reactor vessel intemals guide blocks surround the - ,

sides of the guide lugs is an applicable aging effect. The B&WOG response indicates that the ASME Section XI criteria, which require visual examination of the inner surface of the RV and visual examination of the core guide lug attachment welds, will manage loss of material. The B&WOG proposed to revise BAW-2251 to address loss of material on the sioes of the guide lugs as an applicable aging effect. The staff considers the proposed B&WOG actions acceptable.

(2) Underciad cracking Cracking has been detected under the austenitic stainless steel weld cladding in reactor vessel forgings. These cracks result from intergranular separation of the forging material below the

. austenitic cladding. When cracks are detected, the integrity of the reactor vesselis evaluated as a TLAA to address crack growth.

The impact of underclad cracks on the integrity of the reactor vessel was evaluated by the owners group in topical report BAW-2274. This topical report contained a fracture mechanics analysis that evaluated underclad crack growth, design basis transients and non-design basis transients such as pressurized thermal shock events. The staff evaluation of topical report BAW-

' 2274 is contained in Appendix C of this safety evaluation. The staff found the B&WOG's underciad cracking growth analysis acceptable for the GLRP member plants for the period of extended operation. The staff concluded that neither the design basis transients nor the non-design basis transients will challenge the integrity of the vessel.

The B&WOG stated that treatment of intergranular separations through the TLAA evaluation is 16 l

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consistent with the current licensing basis of the B&W operating plants. For the initial 40-year life of the plants, an evaluation was provided in topical report BAW-10013. BAW-10013 was evaluated by the AEC. Inspection for growth of intergranular separations was not required when

~ the AEC issued the safety evaluation for BAW-10013. According to the B&WOG, the number of design transients that constitute the original design basis would be equivalent to crack growth over 60 years, which would more than envelope the additional 20 years of the period of extended l operation. The B&WOG believes that BAW-2274 satisfies the requirements of Section 54.21(c)

! for an evaluation of the TLAA of intergranular separation and that no additional aging I management programs are needed to manage this TLAA. The staff agrees that the evaluation contained in report BAW-2274 satisfies the requirements of 54.21(c).

According to the ASME Code, flaws larger than the size specified in Table IWB-3510-1 in Section XI of the ASME Code are required to be analyzed to determine whether the reactor vessel is acceptable for service and are required to be reexamined to monitor flaw growth. Altematively, flaws smaller in size than that specified in Table IWB-3510-1 need not be reexamined to monitor flaw growth. The predicted size of an underclad crack would be the sum of the initial flaw size l and the amount of calculated flaw growth. In this case, the maximum initial size reported by I

industry is 0.165 inch in depth and 0.5 inch in length. The calculated amount of flaw growth was only 0.18 inch at the end of the license renewal term. Thus, the predicted flaw size would be

0.345 inch in depth. The predicted size would be less than the value specified in Table IWB-3510-1. I i

The flaw grbwth analysis assumed the underclad crack would grow at a rate equivalent to that of the forging material in reactor water environment. Since the flaws are below the clad, they would not be in contact with the reactor water and would grow at a rate equivalent to that in an air >

environment. The rate of growth in an air environment is less than half the rate in a water i environment for reactor transients. Therefore, the amount of calculated flaw growth would be a conservative value. Since the predicted flaw size is less than the value specified in Table IWB-l 3510-1 and the amount of flaw growth is a conservative value, the staff concludes that additional i

examination of the flaws is not necessary.

1 3.3 Aaina Manaaement Proarams f

l ' As described in Section 2.4, the aging management programs discussed by the B&WOG fall l under 10 CFR 50.60 and 10 CFR 50.61 and the ASME Section XI requirements incorperated in l 10 CFR 50.55a, the plant technical specifications, and licensee commitments in response to NRC l . generic communications. Applicants for license renewal will be responsible for describing any

! such commitments and identifying how such commitments will be controlled.

l 3.3.1 Crackina l

The topical report indicates that ASME Section XI ISI programs for the reactor vessel components within the scope of the report (i.e., those listed in Table 4-1 of the topical report), will l

manage any cracking at welded joints and bolting. The Section XI inservice inspection program  ;

referenced in BAW-2251 is that specified in the 1989 Edition of the ASME Section XI, including '

mandatory Appendices Vil and Vill (1989 Addenda for Appendix Vlli). The 1989 Edition nf the ASME Section XI, including Appendix Vll, has been reviewed and incorporated by reference in

, 10 CFR 50.55a. The staff believes it is essential to include Appendix Vill (Performance 17 '

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f Demonstration Program)in order to provide assurance of the reliability of ultrasonic examinations for the period of extended operation. In view of the above, the staff finds the B&WOG reference to the specific ASME Section XI program acceptable. The staff also concludes that ASME Section XI would manage cracking of the reactor vessel components within the scope of the topic 1 a port for the period of extended operation.

Alloy 600 components in the reactor vessel such as CRDM housings and other penetrations could crack. The B&WOG originally proposed tc use the ASME Section XI program, supplemented by leak detection and surveillance of boric acid, to manage cracking of Alloy 600 components. In an April 1,1997, response to the staff's request for additional information concerning Generic Letter 97-01, " Stress Corrosion Cracking of Control Rod Drive Mechanisms and Other Vessel Head Penetrations," the B&WOG stated: "Each participating plant will address additional requirements for RV head penetrations, including closure head penetrations less than 2 inch N.S. (i.e., thermocouple nozzles at TMl-1 and ONS-2)." Thus, a license renewal applicant referencing the topical report will need to submit its plant-specific program to manage cracking of Alloy 600 components it' '.he reactor vessel in its renewal application for staff review.

Intergranular separations in heat-affected zones (HAZ) of iow carbon steel under austenitic stainless steel weld cladding.is discussed in Section 3.2. For plants that have performed a 40-year flaw growth e ,elysis, the B&WOG has performed an evaluation to extend that analysis to 60 years. This evaluation is discussed in Section 3.4 under TLAA.

3.3.2 Loss of Material Due to Wastaae The topical report indicates that aging management of wastage corrosion and loss of material due to boric acid corrosion is covered by a program based on the licensees' responses to NRC Generic Letter 88-05, " Boric Acid Corrosion of Carbon Steel Reactor Coolant Boundary Components in PWR Plants." The generic letter contains supplemental surveillance programs that have not been incorporated into Section XI. The staff agrees that the individual plant commitments to GL 88-05 will adequately address the wastage effacts of boric acid corrosion of extemal carbon steel reactor vessel components, provided these commitments are incorporated in the plant's FSAR so that changes to them will be controlled by 10 CFR 50.59. Such commitments will provide reasonable assurance that wastage effects will be adequately mitigated during the period of extended operation.

3.3.3 Loss of Material or Closure Inteority for Bolted Closures The topical report indicates that the ASME Section XI ISI program and the plant technical specification leakage detection and leakage limits would manage loss of material and loss of mechanical closure integrity of the reactor vessel pressure-retaining bolted closures. Bolting and flanged surfaces are inspected under ASME Section XI as required by 10 CFR 50.55a. Stress relaxation, according to the topical report, is covered by individual plant's reactor coolant system technical specification leakage limits and system surveillance requirements; any leakage, due to aging, leading to a loss of mechanical closure integrity would be detected and mitigated. If leakage were detected, a root cause evaluation would be performed in accordance with the individual plant's 10 CFR Part 50, Appendix B requirements and corrective measures would be taken. Accordingly, the staff finds these measures acceptable to manage loss of material or closure integrity for bolted closures.

18

4" 3.3.4 Reduction of Fracture Touchness The reduction of fracture toughness of the reactor vessel because of neutron embrittlement is addressed by the reactor vessel materials surveillance program in 10 CFR Part 50, Appendix H, requirements for reactor vessel fracture toughness, upper-shelf energy (USE),

10 CFR Part 50, Appendix G, and requirements regarding pressurized thermal shock (PTS),

10 CFR 50.61(PTS rule). The B&WOG indicates that certain plants have evaluated " thermal shock," which is now addressed by the PTS rule. USE and PTS are evaluated as TLAAs in Section 3.4 below. .

The extent of neutron embrittlement depends on the neutron fluence. As discussed in Section 1.2 above, the B&WOG has submitted a separate Topical Report (BAW-2241) dealing with fast neutron fluence methodologies. Staff review has been completed including review of the responses to two sets of RAls, and there are no outstanding issues. The staff issued a safety evaluation on February 18,1999.

Fracture toughness of the reactor vessel affects the pressure-temperature (P-T) limits in the plant technical specifications. The P-T limits must be updated and submitted for staff review in i accordance with the existing requirements in 10 CFR Part 50, Appendix G for both the current I license term and the renewal term. I 3.3.4.1 Reactor Vessel Materials Surveillance Proaram 10 CFR Part 50, Appendix H, " Reactor Vessel Materials Surveillance Program," includes criteria to monitor changes in the fracture toughness properties of ferritic materials in the reactor beltline region as a result of the exposure of the materials to neutron irradiation and a thermal

' environment. Appendix H requires that the surveillance program design and withdrawal schedule meet the requirements of American Society for Testing Materials (ASTM) '

E185, "Studard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Vessels." in the topical report, the B&WOG references BAW-1543, Revision 4, Supplement 2, " Supplement to the Master Integrated Reactor Vessel Surveillance Program" (MIRVP), to demonstrate continuous management of these aging effects and states that the MIRVP meets the requirements of Appendix H with respect to design and withdrawal schedules for an integrated surveillance program (Appendix H, paragraph lil.C). By letter dated June 11, 1991, the staff determined that the MIRVP (BAW-1543, Revision 3) was an acceptable integrated materials surveillance program. By letter dated July 11,1997, the staff approved the withdrawal schedules for all plants included in BAW-1543, Revision 4, Supplement 2, which includes the plants covered by 5AW-2251. However, the staff approval was for a 40-year license term.

The integrated surveillance program approved by the staff consists of three elements: (a) plant-  !

specific capsules, (b) supplementary weld metal surveillance capsules (SUPCAPS) and (c) high ,

fluence supplementary weld metal surveillance capsules (HUPCAPS). Each licensee l participating in the integrated surveillance program has provided at least six plant-specific capsules to the program. Table I compares the reactor pressure vessel fluence at the end of the .

license renewal term to the highest plant-specific capsule fluence for the Generic License l Renewal Program (GLRP) member plants. There are six SUPCAPS with the target fluence for l the capsules varying from 6.1E18 n/cmr to 1.6E19 n/cm2 . There are eight HUPCAPS with target i fluence varying from 1.3E19 n/cm8 to 2.4E19 n/cm2 . The integrated surveillance program will l provide sufficient data to monitor the effect of radiation on the GLRP member plants. i 19 i

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TABLEI COMPARISON OF RPV FLUENCE AT END OF LICENSE RENEWAL TERM AND HIGHEST CAPSULE FLUENCE FLUENCE AT END HIGHEST PLANT- OF LICENSE CAPSULE RENEWAL TERM FLUENCE

' (E19 nicm2) (E19 n/cm2)

ANO-1 1.44 1.5 OC-1 1.26 1.2 ~

OC-2 1.39 1.2 OC-3 ' 1.35 1.7 TMI-1 1.30 0.817 .

The B&WOG believes that the data obtained during the current term of operation will be valid for the period of extended operation provided the technical basis for the integrated program, as discussed in Chapter 4.0 of BAW-1543, Revision 4, are not violated during the period of extended operation. The B&WOG has identified the following activities that must be addressed on a plant-specific basis at the time of application for license renewal.

1. Neutron fluence at the inside surface of the reactor must be monitored physically or analytically during the period of extended operation.

2. Modifications to the design and operation of the plant that result in changes to the neutron energy spectrum, relative to BAW-1543, must be compared to the energy spectrum in which the capsules were irradiated. If applicable, the current term surveillance data must be adjusted for the subsequent impact on the applicable embrittlement evaluations.

3. Modifications to the design and operation of the plant that result in changes to gamma heating, relative to BAW-1543, must be accounted for in the subsequent impact on the '

applicable embrittlement evaluations for the reactor materials. l l

4. Modifications to the design and operation of the plant that result in changes to the reactor inlet temperature, relative to BAW-1543, must be assessed as to their subsequent impact i on the embrittlement of the reactor materials.

5. Plant-specific limitations must be in place to ensure that the parameters discussed above remain valid during the period of extended operation.

The staff agrees with the B&WOG and finds these individual plant requirements acceptable  !

except: (1) that the neutron fluence must be experimentally monitored by ex-vessel or in-vessel j dosimetry, and (2) if the modifications discussed in items 2, 3, and 4, above occur, the licensee  !

must notify the NRC and propose a program to determine the impact of the modifications.

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. o During the review of BAW-2251, the staff had a question on the need to monitor reactor vessel i I nozzle materials. Appendix H requires reactor vessel components expected to receive a neutron I fluence greater than 1 x 10 " to be subject to a materials surveillance program. The B&WOG 4 estimated that the reactor vessel nozzle attachment welds and the nozzles may be subject to neutron f'uence as high as 1.5 x 10 " at 48 effective full-power years (EFPY). This region of the reactor was not considered to be beltline material for 32 EFPY evaluations. The B&WOG evaluated these welds to determine if they were more limiting than the current beltline welds.

The B&WOG concluded that these welds were not limiting materials and that they were not subject to surveillance. The amount of embrittlement of beltline materials that are not in surveillance programs is determined using the methodology in RG 1.99, Rev. 2. The amount of ,

embrittlement in the surveillance materials is compared to the amount of embrittlement predicted i by RG 1.99, Rev. 2 to determine whether the embrittlement for the beltline non-surveillance materials is adequate. Therefore, the staff finds the B&WOG determination acceptable. l 3.4 Time-Limited Aoino Analyses Time-limited aging analyses are defined in 10 CFR 54.3 as those licensee calculations and analyses that:

(1) involve systems, structures, and components within the scope of license renewal, as stated in 10 CFR 50.54(a); i l

(2) consider the effects of aging -

l (3) involve time-limited assumptions defined by tha current operating term, for example,40 l years; j l

(4) were determined to be relevant in making a safety determination; (5) involve conclusions or provide the bases for conclusions related to the capability of the system, structure or component to perform its intended functions, as stated in 10 CFR 50.54(b); and (6) are contained or incorporated by reference in the current licensing basis.

Section 54.21(c)(1) requires the applicant to demonstrate that:

l (i) the analyses remain valid for the period of extended cperation; (ii) the analyses have been projected to the end of the period of extended operation; or (iii) the effects of aging on the intended functions (s) will be adequately managed for the period of extended operation.

The TLAAs evaluated in BAW-2251 for the reactor vessel involve:

(1) fatigue of metallic components; (2) analyses and calculations performed to show compliance with NRC regulations (10 CFR r

50.60 and 50.61) concerning reduced fracture toughness of reactor vessel materials (PTS 21

'r and reduced Charpy USE); and

_ (3) growth of intergranular separations in low alloy steel forging heat-affected zones under stainless steel weld deposit cladding (underclad cracking).

3.4.1 Fatiaue (includina Environmentally Assisted Fatioue)

Section 4.5.1 of BAW-2251 describes the B&WOG evaluation of metal fatigue of the reactor vessel components as a TLAA for license renewal. According to the B8iWOG, the fitness of the reactor vessel for specified operating conditions involving cyclic application of loads and thermal conditions was determined in accordance with the requirements in Subsection A, Article 4, Paragraphs N-415 and N-416, of the 1965 Edition of the ASME Boiler and Pressure Code,

'Section ill, with Addenda through Summer 1967. In addition, the core flood nozzle venturis were designed for cyclic operation in accordance with the requirements in Subarticle NB-3200 of the 1971 Edition of Section Ill. The design cyclic loadings are defined in the component denign specification. These design cyclic loadings were used to calculate the ability of the components to withstand cyclic operation without fatigue failure. The ability to withstand these design cyclic loadings without fatigue failure is expressed in terms of a fatigue usage factor as defined in the ASME Code.

' On the basis of its review of the final safety analysis reports (FSARs) for Arkansas Nuclear One Unit 1. Three Mile Island Unit 1, and the Oconee Nuclear Station, the B&WOG determined that the Class 1 components were designed for a cyclic service life of 40 years. In order to demonstrate that the fatigue usage factors remain valid through the extended period of operation, the B&WOG reviewed component design basis documentation and compiled data on the design transients used to calculate fatigue usage factors. The B&WOG identified the controlling design transients from the usage factor calculations for the participating utilities.

According to B&WOG, all of its participating utilities monitor occurrences of design transients at their plants. B&WOG projected the number of these transients for 60 years of plant operation based on the monitoring of design transients by its partidpating utilities. According to B&WOG, i the number of transients projected for 60 years of plant operation is less than the number of cycles : assumed in the original design. On this basis, B&WOG concluded that the existing usage factor calculations would remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).-

l The staff, in an RAI dated December 19,1996, requested that B&WOG describe the methods  !

and instrumentation that each of the participating utilities use to monitor design transients.

B&WOG, in its April 1,1997, RAI response, indicated that the specific plant procedures and '

instrumentation used to assess design transients would be addressed on a plant-specific basis at the time of application for license renewal. Projecting the number of transients during a plant's operational history of 60 years is an acceptable method to demonstrate that the requirements of 10 CFR 54.21(c)(1)(i) have been met. Thus, a license renewal applicant needs to provide the plant-specific procedures and instrumentation used to assess the number of operational I transients in its renewal application for staff review.

In the report, B&WOG ' stated that the fatigue usage factor for the Oconee Nuclear Station reactor vessel studs is projected to exceed the ASME Code allowable usage factor at the end of the period of extended operation. B&WOG further stated that this issue would be addressed in any license renewal application submitted for Oconee. The staff, in an RAI dated December 19,1996 22

.~ ..

• requested that B&WOG provide additional information regarding the calculated usage factor for the Oconee reactor vessel studs. The staff concern was whether the usage factors for reactor vessel studs at the other B&W facilities would also be expected to exceed the ASME Code allowable limit at the end of the period of extended operation. B&WOG, in its April 1,1997, RAI response, provided the reasons that the usage factor calculated for the Oconee reactor vessel studs was higher than for the other participating facilities. First, the number of design heatup and cooldown cycles had been increased from 240 to 360 for the Oconee plants for the current j license term. The other plants still used 240 cycles for design. Second, the licensee performed an evaluation of the effects of a reactor coolant pump restart with voids in the system far the Oconee plants. The licensee assumed 35 occurrences of this transient over the lifetime of each plant. This transient is not in the design specification of the other plants. The staff will review fatigue of the reactor vessel studs at Oconee en a plant-specific basis. In addition, as stated above, the staff will review the plant-specific procedures and instrumentation used to assess the number of design transients when the license renewal application is submitted. The staff review will include the number of operating cycles applicable to the reactor vessel studs.

! Section 4.5.1 of the topical report also contains a discussion of environmentally assisted fatigue of metal components. Current test data indicate that the design fatigue curves of the ASME Code may not be conservative for nuclear power plant primary system environments. The ASME fatigue curves were developed from laboratory specimens tested in air at room temperature. The a current test' data indicates there could be a significant reduction in the fatigue life of metal I components in a reactor primary system environment. The staff addressed the issue of environmentally assisted fatigue in Generic Safety issue (GSI) 166, " Adequacy of Fatigue Life of l Metal Components." The staff recommendations are contained in SECY-95-245. In SECY-95-245, the staff did not recommend backfit of new environmental fatigue curves to operating plants.- This recommendation was based, in part, on conservatism j identified in the existing fatigue analyses and on a risk assessment considering a 40-year plant j design life.

l The staff has not resolved the issue regarding environmental fatigue data for license renewal. A further assessment is currently being performed under GSI-190, " Fatigue Evaluation of Metal l Components for 60-Year Plant Life." in SECY-95-245, the staff indicated that it would consider

whether license renewal applicants need to evaluate a sample of components with high-fatigue l usage factors, using the latest available environmental fatigue data. The staff further indicated l that if the generic safety issue has not been resolved before the issuance of a renewed license, the applicant should submit its technical rationale for concluding the effects of fatigue are

adequately managed for the extended period or until the resolution of the generic safety issue becomes available (60 FR 22484, May 8,1995). The evaluation of a sample of components with l- high fatigue usage factors using the latest available environmental fatigue data will provide an i acceptable technical rationale to address GSI-190.

l B&WOG indicated that reanalysis to assess environmentally assisted fatigue, as recommended in SECY 95-245, is required only if the fatigue usage factor exceeds the ASME Code limit (calculated using the current ASME fatigue curves) during the extended period of operation. In the December 19,1996, RAI, the staff indicated that this B&WOG interpretation was incorrect.

The staff intended that those components that had high usage factors (calculated with fatigue curves that included environmental effects) be evaluated further for any extended period of operation. In its April 1,1997, RAl response, B&WOG indicated that a generic assessment of environmentally assisted fatigue was completed for the reactor vessel using environmental factors based on the model described in NUREG/CR-6335 (Reference 7). The factors were 23

e Q , O.

r applied to the items studied in NUREG/CR-6260 (Reference 8). According to B&WOG, all reactor vessel items were found acceptable for the extended period of operation after applying the appropriate environmental factors.-

Additional testing of stainless steel components, performed after publication of NUREG/CR-6335, indicates that the reduction in fatigue strength for stainless steel components in low- oxygen primary system environments may be greater than previously identified. The staff is currently assessing the significance of this new stainless steel data. Therefore, the stainless steel components identified in NUREG-6260 may require additional evaluations using the latest I stainless steel test data.- High usage factors for the B&W reactor vessel involved low alloy steel I and Alloy 600 components. Since no stainless steel components with high usage factors were identified in NUREG 6260, the new stainless steel test data should not impact B&WOG's assessment described above for the reactor vessel.-

On the basis of the information described above, the staff considers B&WOG's assessment of the impact of environmental fatigue data on the reactor vessel provides an adequate basis to demonstrate that fatigue will be managed during the extended period of operation. Accordingly, the staff finds the TI.AA evaluation performed by the B&WOG on fatigue of the reactor vessel I components for the period of extended operation acceptable for the GLRP member plants except for Oconee reactor vessel studs, which the staff will evaluate on a plant specific basis. The staff also finds that the B&WOG has adequately addressed GSI-190 regarding environmentally assisted fatigue of the reactor vessel components for the GLRP member plants for license renewal.

3.4.2 Reduction of Fracture Touahness

)

The regulations goveming reactor vessel integrity are in 10 CFR Part 50:

(4) Section 50.60 requires all light-water reactors to meet the fracture toughness and material surveillance program requirements for the reactor coolant boundary set forth in ,

Appendices G and H to 10 CFR Part 50; and  !

(2) Section 50.61 contains fracture toughness requirements for protection against PTS l events. 1 Regulatory Guide (RG) 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials,"

describes general procedures acceptable to the staff for calculating the effects of neutron radiation embrittlement of the low alloy steels currently used for light-water-cooled reactor vessels. The fracture toughness of the reactor coolant pressure boundary required by 10 CFR Past 50 is necessary to provide adequate margins of safety during any condition of normal plant operation, including anticipated operational occurrences and system hydrostatic tests.

The B&WOG evaluated analyses of Charpy USE (Appendix G to 10 CFR Part 50) and PTS (10 CFR 50.61) as TLAAs for license renewal. The B&WOG evaluated the reactor vessel surveillance program (Appendix H to 10 CFR Part 50) as part of its aging management review, and the staff evaluation of the surveillance program is contained in Section 3.4.4.1 above.

During the review of the topical report, the staff had a question regarding the need to update the 24

reactor vessel fracture toughness estimates with new data as they become available. In its August 11,1997, RAI response, the B&WOG states: "Each license renewal applicant will define a process to ensure that the time-dependent parameters used in the TLAA evaluations reported in BAW-2251 are tracked so that the TLAA remains valid through the period of extended <

operation. The process will be defined on a plant-specific basis at the time of the licensee renewal application." This is acceptable to the staff. A license renewal applicant needs to i

describe such a process in its application for staff review. If new information affects the '

conclusions of the topical report for the applicant's plant, the. applicant needs to update its TLAA evaluations as appropriate and provide the updated evaluations in its renewal application for staff review.

Embrittlement of the reactor vessel will be managed to ensure intended functions of the reactor vessel for 60 years. For the staff to determine if the plant could be operated for 60 years, an applicant must show that there will be available operating window between the pressure-temperature limits and the net positive suction curves for the RC pumps for 60 years.

3.4.2.1 Charov Upoer-Shelf Enerav Appendix G of 10 CFR Part 50 requires that " reactor vessel beltline materials must have Charpy upper-shelf energy . . of no less than 75 ft-Ib initially and must maintain Charpy upper-shelf-energy throughout the life of the vessel of no less than 50 ft-lb. unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation, that lower values of Charpy upper-shelf energy will provide margins of safety against fracture equivalent to those  ;

required by Appendix G of Section XI of the ASME Code." The issue of low upper-shelf fracture  !

toughness for Linde 80 welds in B&W vessels was addressed using fracture mechanics analysis

' by the B&WOG for its 16 member plants for a period of 32 EFPY in topical reports BAW-2192PA j and BAW-2178PA. Both reports were apprdVed by NRC on March 29,1994, as in accordance  !

with the methodology and criteria contained in the ASME Code Case N-512, which was adopted )

later as Appendix K of the ASME Code. This effort was related to Generic Letter (GL) 92-01, Revision 1, " Reactor Vecsel Structural integrity," for demonstrating that although the predicted end-of-license (EOL) USE is != tow 50 ft-lb for some Linde 80 welds in B&WOG vessels, these welds will still provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code.  !

Topical report BAW-2251 includes another topical report, BAW-2275 (Reference 9), which addresses the issue of low USE for Linde 80 welds for license renewal, i.e.,for a period of 48 EFPY. The staff used the calculation procedures and evaluation criteria in Appendix K of Section XI of the ASME Code to conduct the review.

Appendix B below contains the staff's review of BAW-2275, which concludes that the B&WOG's analytical results satisfy the acceptance criteria of Appendix K of the ASME Code. Hence, the Linde 80 welds of the five plants addressed in the topical report have margins equivalent to those i of Appendix G of Section XI of the ASME Code. The staff has also examined the recent best-

- estimate chemistry data from Framatome Technologies inspection Report No. 99901300/97-01 (Reference 11), and concluded, as explained in Appendix B, that the recent data has no impact i on the results and conclusions made in this evaluation. In summary, the staff finds the B&WOG's .

evaluation of the Charpy USE acceptable for the GLRP member plants for the period of extended operation. Details of the staff's evaluation are provided in Appendix B.

25

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- 3.4.2.2 Pressurized Thermal Shock 10 CFR 50.61 provides requirements for protection against PTS transients in pressurized-water reactors (PWRs). This regulation requires licensees to perform an assessment of the projected values of a reference temperature, RTpTs for the end-of-life fracture toughness of all reactor beltlinq material. If the projected reference temperature exceeds the screening criteria established in 10 CFR 50.61, the licensee is required to implement such flux reduction programs as are reasonably practicable to avoid exceeding the screening criteria. The schedule for implementation of such programs may take into account the schedule and anticipated approval by the Director, NRR, of detailed plant-specific analyses submitted to demonstrate acceptable risk with RTpTs above the screening limit. If the licensee cannot avoid exceeding the screening criteria by using a flux reduction program, it must submit a safety analysis to determine what actions are necessary to prevent potential failure of the reactor vesse!. . Section 50.61 also permits the licensee to perform a thermal annealing treatment to recover fracture toughness, subject to the requirements of 10 CFR 50.66. The regulations require updatirig of the pressurized thermal shock assessment upon a request for a change in the expiration date of operation of a facility. Therefore, the RTns value must be calculated for the reactor life extensi,on period of 48 effective full-power years. RTns is the RTum at the end of life.

The. screening criterion established by 10 CFR 50.61 is 270

• F for plates, forgings, and axial welds. The screening criterion is 300

• F for circumferential welds. According to this regulation, if the calculated RTns for the limiting reactor beltline mateJals is less than the specified screening criteria, the vesselis acceptable with regard to the risk of vessel failure during pressurized thermal shock transients.

Regulatory Guide (RG) 1.99, Revision 2, specifies two methodologies for determining the effect of kradiation on RTuor. The first methodology, Position 1.1, does not rely on plent-specific surveillance data to calculate delta RTsm (i.e., the mean value of the adjustment or shift in reference temperature caused by irradiation). The delta RTum is determined by multiplying the chemistry factor from the tables in RG 1.99 by the fluence factor. Similarly, the fluence factor is calculated from the neutron flux using an equation or a figure in RG 1.99.

The second methodology in RG.1.99, Revision 2, Position 2.1, relies on plant-specific surveillance data to determine the delta RTum. In this methodology, two or more sets of surveillance data are needed. Surveillance data consists of a measured delta RTum for a corresponding fluence. The fluence is converted to a fluence factor using an equation provided in RG 1.99. RG 1.99 specifies a procedure and criteria for determining whether the surveillance data are credible. Using a ratio procedure specified in RG 1.99, Position 2, the measured delta RTum values are normalized to the best-estimate chemical composition of the vessel weld. A best-fit line is then determined relating the adjusted delta RTum values to the fluence factor. This best-fit line has a zero y-intercept. Therefore, delta RTug will be zero at a fluence factor equal to zero. The slope of the best-fit line will equal the chemistry factoi. The scatter around the best-fit line, that is, the difference in the predicted value and the measured value for delta RTum, must be less than 28'F for weld metal for the surveillance data to be defined as credible. When a credible surveillance data set exists, the chemistry factor determined from the surveillance data can be used in f.eu of the values in the table in Regulatory Guide 1.99, Revision 2, and the sigma de;ta values used in the determination of the margin term can be reduced to one-half of 28'F for welds.

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e The B&WOG originally proposed to use the " master curve" methodology to estimate the initial RTuor values. However, in its August 11,1997, RAI response, the B&WOG withdrew the use of

' that methodology and decided to use the currently accepted values. Attached to the B&WOG letter of August 11,1997, were the revised calculations presented in the Framatome Technologies Calculation Summary Sheet entitled *B&W 177-FA Reactor Vessel Fracture Toughness Properties". (Document identifier 32-1240132-03). Wdh reference to the recalculations, the B&WOG has stated that its calculated values of RTns for all beltline materials at ANO-1, ONS-1, and ONS-3 are below the applicable PTS screening criteria at 48 EFPY, Using Position 1.1 of RG 1.99, Revision 2, the calculated value of RTpr for weld WF-25 at ONS-2 exceeds screening criteria at 45 EFPY and the longitudinal weld at TMI-1 exceeds the screening criteria at 32.1 EFPY Position 2 of RG 1.99, Revision 2, does not apply to ONS-2 and TMI-1 because the 32-1240132-03 report states: " Surveillance data credibility in question." The staff has reviewed the calculations stated to be derived from the " table values" in 10 CFR 50.61 and found them to be conservative and, therefore, acceptable.

The following table provides the most limiting RTns calculation for each of the plants as calculated by the B&WOG: )

1 MOST LIMITING RTpre CALCULATIONS PLANT RV Beltline Weld Material Initial RTns,'F Screening Location identification RCTuor,

• F 48 EFPY Criteria ANO 1 US to LS Cire. WF-112 -5 278 300 Weld (ID 100%)

Oconee 1: IS Longit. Weld SA-1073 -5 228 270 (Both 100%)

Oconee 2 US to LS Cire. WF-25 +7 304 300 Weld (ID 100%)

Oconee 3 NB to US Cir. WF-200 -5 250 300 Weld (100%)

TMl1 LS Long. Weld SA-1526 -7 296 270 (37% OD)

The staff finds the B&WOG evaluation of PTS acceptable. Except for ONS-2 and TMI-1, the GLRP member plants may reference BAW-2251 in their renewal applications to demonstrate l compliance with 10 CFR 50.61 for the period of extended operatien. In its August 11,1997, RAI j response, the B&WOG indicates that ONS-2 and TMI-1 will provide updated predictions of RTns l for limiting welds WF-25 and SA-1526, respectively, at the time of plant-specific application for  !

license renewal. For plants whose RTns value for 48 EFPY exceeds the corresponding PTS screening criterion, a license renewal applicant must address the requirements in 10 CFR 50.61.

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3.4.3 Interaranular Seoarations Under Weld Claddina

)

As discussed in Appendix C, the B&WOG has previously performed flaw growth analysis of underclad cracks in B&W reactor vessels based on 40 years of plant operation. Topical report BAW-2251 references another topical report, BAW-2274 (Reference 10), which contains the l

B&WOG's evaluation of underclad cracks beneath austenitic stainless steel cladding for the license renewal term. B&WOG's methodology in performing the flaw evaluation is consistent with

! the current well-established flaw evaluation procedure and criteria in the ASME Code and, therefore, is adequate. The staff examined the unique aspects of the B&WOG approach and found them adequate also. The B&WOG's methodology, as discussed in Appendix C, includes the following conservatisms: (1) using the maximum crack depth of 0.165 inch reported by the industry as the initial crack depth instead of the depth of 0.10 inch reported on the B&W RPVs; (2) assuming all underclad cracks are surface cracks; (3) using the fatigue crack growth rate for

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surface flaws in water reactor environment; and (4) using a safety factor 17% more than that

. specified by the ASME Code for Level A and B (normal and upset) loading and 72% more than for Level C and D (emergency and faulted) loading. Additional details of the staff evaluation are provided in Appendix C below. In summary, the staff finds the B&WOG's underclad cracking flaw growth analysis acceptable for the GLRP member plants for the period of extended operation.

4.0 CONCLUSION

S The staif has reviewed the subject B&WOG topical report (Reference 2) and additional informaiion submitted by the B&WOG. On the basis of its review, the staff concludes that the B&WOG topical report provides an acceptable demonstration that the aging effects of reactor vessel components within the scope of this topical report will be adequately managed for the GLRP member plants with the exception of the noted renewal applicant action items, so that there is reasonable assurance that the reactor vessel components will perform their intended functions in accordance with the CLB. The staff also concludes that, upon completion of the renewal applicant action itema set forth in Section 4.1 below, the B&WOG topical report provides an acceptable evaluation of time-limited aging analyses for the reactor vessels for the GLRP member plants for the period of extended operation.

Any B&WOG GLRP member plant may reference this topical report in a license renewal application to satisfy the requirements of (1) 10 CFR 54.21(a)(3) for demonstrating that the effects of aging on the reactor vessel components within the scope of this topical report will be l adequately managed and (2) 10 CFR 54.21(c)(1) for demonstrating that appropriate findings be )

made regarding evaluation of time-limited aging analyses for the reactor vessels for the period of i extended operation. The staff also concludes that, upon completion of the renewal applicant i action items set forth in Section 4.1 below, referencing this topical report in a license rer.ewal  !

application and summarizing in an FSAR supplement the aging management programs and the TLAA evaluations contained in this topical report will provide the staff with sufficient information to l

make the necessary findings required by Sections 54.29(a)(1) and (a)(2) for components within ,

the scope of this topical report. j l

4.1 Renewa! Aoolicant Action items -j i

. The following are license renewal applicant action items to be addressed in the plant-specific ,

license renewal application when incorporating the B&WOG topical report in a renewal j 28 i

E application:

(1) The license renewal applicant is to verify that its plant is bounded by the topical report.

Further, the renewal applicant is to commit to programs described as necessary in the topical report to manage the effects of aging during the period of extended operation on the functionality of the reactor vessel components. Applicants for license renewal will be responsible for describing any such commitments and identifying how such commitments ]

will be controlled. Any deviations from the aging management programs within this topical i report described as necessary to manage the effects of aging during the period of extended operation and to maintain the functionality of the reactor vessel components or other information presented in the report, such as materials of construction, will have to be identified by the renewal applicant and evaluated on a plant specific basis in accordance with 10 CFR 54.21(a)(3) and (c)(1).

(2) A summary description of the programs and evaluation of TLAAs is to be provided in the license renewal FSAR supplement in accordance with 10 CFR 54.21(d).

(3) Since the staff has not made any finding on whether the B&WOG topical report provides the complete list of reactor vessel components subject to an aging management review or whether the scoping methodology is adequate, individual plant applicants will need to provide a comprehensive list of structures and components subject to an aging management review and the methodology for developing this list as part of their license renewal applications. Any components determined by the applicant to be subject to an aging management review for license renewal but not withira the scope of the topical report are required to be addressed in the license renewal application.

(4) The B&WOG has determined that the lower CRDM service support structure, including the weld that connects the lower CRDM service support skirt to the reactor vessel closure head, and the reactor vessel support skirt, including the weld that connects the reactor ,

vessel support skirt to the transition forging, are subject to an aging manageent review I for license renewal. However, the B&WOG has decided to exclude them from the scope l of the topical report. Thus, a renewal applicant needs to address them in its license '

renewal application.

l (5) The license renewal application for Oconee needs to address the fatigue evaluation of the reactor vessel studs on a plant-specific basis.

(6) A license renewal applicant needs to discuss the plant-specific methodology and instrumentation used to assess the number of operat.onal transients in its renewal application for staff review. The staff review will also include the number of operating cycles applicable to the reactor vessel studs.

(7) The B&WOG identifies flaw growth acceptance in accordance with the ASME Section XI ISI program as a TLAA, but indicates that flaw growth acceptance evaluation is plant-

specific, is not within the scope of the report, and will be resolved on a plant-sp?cific basis. Thus, a license renewal applicant neecs to address it in the renewal application.

29

(8) Alloy 600 components in the reactor vessel such as CRDM housings and other penetrations may be subject to crack initiation and growth. The B&WOG originally proposed to use the ASME Section XI program, supplemented by leak detection and surveillance of boric acid, to manage cracking of Alloy 600 components. In an April 1, 1997, response to the staffs request for additional information conceming Generic Letter 97-01, " Stress Corrosion Cracking of Control Rod Drive Mechanisms and Other Vessel Head Penetrations," the B&WOG stated: *Each participating plant will address additional requirements for RV head penetrations, including closure head penetrations less than 2 inch N.S. (i.e., thermocouple nozzles at TMl-1 and ONS-2)." Thus, a license renewal applicant referencing the topical report will need to submit its plant-specific program to manage cracking of Alloy 600 components in the reactor vessel in its renewal application for staff review.

(9) During the review of the topical report, the staff had a question regarding the need to update the reactor vessel fracture toughness estimates with new data as it become available. In its August 11,1997, RAI response, the B&WOG states: *Each license renewal applicant will define a process to ensure that the time-dependent parameters used in the TLAA evaluations reported in BAW-2251 are tracked such that the TLAA remains valid through the period of extended operation. The process will be defined on a plant-specific basis at the time of the licensee renewal application." Thus, a license renewal applicant needs to describe such a process in its application for staff review. If new information affects the conclusions of the topical report for the applicant's plant, the applicant needs to update its TLAA evhluations as appropriate and provide the updated evaluations in its renewal application for staff review.

(10) in its August 11,1997, RAI response, the B&WOG indicated that Oconee Unit 2 and TMl Unit 1 will provide updated predictions of RTers for welds WF-25 and SA-1526, respectively, when the plant specific application for license renewal is submitted. For plants with an RTpts value for 48 EFPY exceeding the corresponding PTS screening criterion, a license renewal applicant must address the requirements in 10 CFR 50.61(b)(3) by developing, and requesting staff approval for reasonably practicable flux reduction programs to avoid exceeding.the PTS criterion.

(11) If an applicant has installed flow stabilizers using Alloy 600 and/or Alloy 82/182 weld material, the applicant must include the flow stabilizers in its Alloy 600 aging management program. Alloy 600 and Alloy 82/182 weld materials are susceptible to cracking in primary water environments.

(12) Embrittlement of the reactor vessel will be managed to ensure intended functions of the reactor vessel for 60 years. For the staff to determine if the plant could be operated for 60 yeam, an applicant must show that an operating window will be available between the pressure-temperature limits and the net positive suction curves for the RC pumps for 60 years. Otherwise, the applicant will propose aging management activities to minimize the extent of embrittlement, or other attematives, to permit safe plant operation for 60 years.

- Should the applicant show that the reactor could only be operated for a time period less than 60 years, the duration of the renewed license, if granted, would be limited to that time period.

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r (13) The neutron fluence must be experimentally monitored by ex-vessel or in-vessel dosimetry, and if modifications to the design and operation of the plant changes either the neutron energy spectrum, gamma heating or the reactor inlet temperature, as discussed in section 3.3.4.1 of this safety evaluation, the licensee must notify the NRC and propose a program to determine the impact of tne modifications. f j

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Primary Contributors: B. Elliot, H. Conrad, S. Sheng, J. Fair, and M. Razzaque l '

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5.0 REFERENCES

1. Part 54, "Reouirements for Renewal of Operating Licenses for Nuclear Power Plants,"

Federal Reoister, Vol. 60, No. 88, pp. 22461-22495, May 8,1995.

2. BAW-2251, " Demonstration of the Management of Aging Effects for the Reactor Vessel,"

Bsbcock & Wilcox Owners Group, June 1996.

3. BAW-2241P, " Fluence and Uncertainty Methodologies," Babcock &.Wilcox Owners Group, April 1997.

4. Boiler and Pressure Vessel Code,Section XI,

• Rules for Inservice Inspection of Nuclear Power Plant Components," the American Society of Mechanical Engineers,1989.

5. BAW-2243A, " Demonstration of the Management of Aging Effects for the Reactor Coolant System Piping," Babcock & Wilcox Owners Group, June 1996.

6. " Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants," Working Draft, September 1997.

7. NUREG/CR-6335," Fatigue Strain-Life Behavior of Carbon and Low-Alloy Ferritic, Austenitic Stainless Steels, and Alloy 600 in LWR Environments," August 1995. ,

8. NUREG/CR-6260," Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," March 1995.

l

9. BAW-2275,

• Low Upper-Shelf Toughness Fracture Mechanics Analysis of B&W Designed Reactor Vessels for 48 EFPY," Babcock & Wilcox Owners Group, August 1996.

10. BAW-2274, " Fracture Mechanics Analysis of Postulated Underclad Cracks in B&W Designed Reactor Vessels for the Period of Extended Operation," Babcock & Wilcox Owners Group, December 1996.

11. Letter from S. C. Black, NRC, to E. R. Kane, Framatome, dated January 28,1998.

I 32 L

i

. , ' , d APPENDIX A LIST OF CORRESPONDENCE

1. Letter to U.S. NRC, attention B. K. Grimes, from D. K. Croneberger of B&WOG Generic License Renewal Programs, dated June 27,1996, submitting BAW-2251, " Demonstration of the Management of Aging Effects for The Reactor Vessel," June 1966.

2. Letter to D. K. Croneberger of B&WOG Generic License Renewal Programs from P. T. Kuo'of NRC, dated August 28,1998, transmitting Request for Additional information (RAls) 1 through 7.

3. Letter to D. K. Croneberger of B&WOG Generic License Renewal Programs from P. T. Kuo of NRC, dated December 19,1996, transmitting RAls 8 through 17.

.:. Letter to D. J. Firth of B&WOG Generic License Renewal Programs from P. T. Kuo of NRC, dated April 2,1997, transmitting RAls 18 through 26.

5. Letter to U.S. NRC, attention M. Slosson, from D. J. Firth of B&WOG,.dsted June 20,1997, providing responses to RAls 18 and 19.

6. NRC Meeting Minutes by R. J. Prato, dated July 1997;

Subject:

Summary of Meeting Between the U.S. NRC and B&WOG to Discuss the Request for Additional information

, Related to the B&W Reactor Vessel Topical Report.

7. Letter to U.S. NRC, attention S. F. Newberry, from D. K. Croneberger of B&WOG, dated July 31,1996,

Subject:

NRC Review of B&WOG Topical Report BQW-2251.

8. Letter to U.S. NRC, attention R. K. Anand, from D. J. Firth of B&WOG, dated October 21, 1997, submitting a Comparison of B&WOG Measurement-Based and Calculation-Based Fluence Methodologies.

F 33 1

APPENDIX B ASSESSMENT OF TOPICAL REPORT BAW-2275

" Low Upper-Shelf Toughness / Fracture Mechanics Analysis of B&W Designed Reactor Vessels for 48 EFPY" Introduction Topical Report BAW-2275 (Reference B.1) addresses the issue of low upper-shelf fracture toughness for Linde 80 welds in Babcock & Wilcox (B&W) vessels. The B&WOG addressed this issue using fracture mechanics analysis for its 16 member plants for a period of 32 effective full power years (EFPY) in topical reports BAW-2192PA and BAW-2178PA. The NRC approved both reports on March 29,1994 (References B.2 and B.3), using the methodology and criteria contained in American Society of Mechanical Engineers (ASME) Code Case N-512, which was adopted later as Appendix K in the ASME Code. These topical reports were in response to Generic Letter (GL) 92-01, Revision 1. These topical reports demonstrated that although the predicted end-of-license (EOL) upper-shelf energy (USE) is below 50 ft-lb for some Linde 80 welds in B&WOG vessels, these welds will still provide margins of safety against fracture l

equivalent to those required by Appendix G of Section XI of the ASME Code. l The current topical report addresses the same icsue of low USE for Linde 80 welds, but for an extended license period of 48 EFPY. The staff used the calculation procedures and evaluation criteria in Appendix K of the ASME Code to conduct the review. Appendix K was developed from Code Case N-512 and incorporated into the ASME Code as an appendix in 1995. The plants ,

participating in the current effort include only five of the 16 plants covered by BAW-2192PA and BAW-2178PA,namely the following: Oconee, Units 1, 2, and 3 (Oconee-1, -2, and -3), Three  :

Mile Island, Unit 1 (TMI-1), and Arkansas Nuclear One, Unit 1 (ANO-1). l The B&WOG's Assessment on Low Uooer-Shelf Touahness The B&WOG performed fracture mechanics analyses according to the acceptance criteria and evaluation procedures of Appendix K of the ASME Code. In these fracture mechanics analyses, the crack-resistance parameter for ductile fracture, J, is the J-integral due to the applied loads, and J, is the J-integral fracture resistance for the material based on the deformation theory of plasticity. For Level A and B loading, the B&WOG used a postulated 6:1 semi-elliptical surface flaw with a depth of 1/4T (T = reactor pressure vessel wall thickness), and safety factors of 1.15 for the pressure and 1.0 for the 100*F/ hour cooldown transient to satisfy the first acceptance .

criterion of J < Joi, where Jo, is J, at a ductile flaw extension of 0.1 inch. For these same l loadings and postulated surface flaws, the B&WOG used safety factors of 1.25 for the pressure and 1.0 for the cooldown transient to satisfy the second acceptance (stability) criterion of dJ/da < l dJ /da at J =J , where the parameter a is the crack depth. For Level C loading, the B&WOG used a postulated 6:1 semi-elliptical surface flaw with a depth of the clad plus 1/10T and a safety factor of 1.0 for the most severe transient for Level C and D loadings (i.e., Hot Leg LOCA) to satisfy both the acceptance criteria of J < Jo., and dJ/da < dJ /da at J = J,. For Level D loading, i four transients, including Hot Leg LOCA, have been analyzed. The B&WOG used the same postulated surface flaw and loading safety factors that were employed in the Level C analysis. A stability criterion identical to the second criterion for Level C loading was used in the Level D 34

7 o , . O i analysis, and the best-estimate J, curve was used rather than the lower-bound (conservative) J, curve used in the Level A, B, and C analyses. The additional criterion in the Level D analysis of limiting the stable flaw extension to 75% of the RPV wall thickness was also considered by the B&WOG. For the Level A and B analyses, the B&WOG used the evaluation procedures specified in Appendix K of the ASME Code. For Level C and D analyses, the evaluation procedures have not been specified in Appendix K. The B&WOG used a methodology based on a one-dimensional finite element model for the thermal stress analysis and linear elastic fracture mechanics (LEFM) for the subsequent fracture analysis.

~

The B&WOG tabulated its results for Level A and B loading in Table 5-3, and demonstrated that the limiting weld of all five plants meets the first criterion. Among the five, Weld SA-1526 of TMI-1 is most hmiting, with a Jo,/J ratio equal to 1.09. Figure 5-1 further indicates that Weld SA-1526 also satisfies the second criterion. For Level C loading, the B&WOG tabulated its results in Table 6-1, and demonstrated that i se limiting weld of all five plants meets the first criterion.

Again, Weld SA-1526 of TMI-1 is most limiting, with a Jo,/J ratio equal to 3.26. Figure 6-6 shows that for weld SA-1526, the intersection point of applied J and J,is below J., the toughness for ductile fracture. This indicates no onset of ductile tearing; therefore, the stability criterion for Level C and D loading and the limited extension criteria for Level D loading are automatically satisfied.

The Staff's Assessment on Low Upper-Shelf Touchness The B&WOG's methodology in performing the equivalent margins analysis of the low USE issue for Level A and Level B loading is consistent with the Appendix K criteria and evaluation procedures in the ASME Code, and is, therefore, adequate. For Level C and D analyses, for which Appendix K does not specify a detailed evaluation procedure, the B&WOG used an approach consistent with the methodology in topical report BAW-2178PA. Based on oast experience in reviewing the low USE issue at 32 EFPY (Reference B.3) and recent information on chemistry data for Linde 80 welds, the staff decided to examine the following two technical issues. The first needs to be reviewed because the previous SEs did not evaluate it; the second needs to be re-evaluated because this issue is sensitive to higher fluence (EFPYs).

The Impact of Recent Best-estimate Chemistry Data from Framatome The NRC performed an inspection at Framatome Technologies, Inc., on May 19-21,1997, to review records pertaining to the chemical composition of automatic submerged-arc welds in RPVs fabricated by B&W. Since then, Framatome has compiled all chemistry data, established traceability, and assessed the data. On June 30,1997, Framatome sent to the B&WOG and NRC the best-estimate chemistry data for Linde 80 welds (Reference B.4). The current topical report was based on the best-estimate data before mid-1997; therefpre, the impact of the most recent chemistry data of B&WOG RPV welds on the conclusions of the current report has to be evaluated. The staff has examined the recent best-estimate chemistry based on total population and coil-to-coil classification by Framatome, and determined that the maximum increase in best-estimate copper due to this revision is 0.01% for B&W fabricated welds. According to the chemistry table of Regulatory Guide (RG) 1.99, Revision 2, this small increase in best-estimate copper would only slightly reduce the fracture toughness. The staff examined additional margins for all plants after they have met the Appendix K acceptance criteria, and concluded that the recent Framatome data have a negligible impact on the conclusions of the topical report, i.e.,

35

.- , .e acceptable margins remain for all plants after accounting for the reduction in fracture toughness.

The imolication of usina the B&WOG Fluence-Cooner J Model Unlike the Eason J, models in NUREG/OR-5729, which were derived from a larger material database and had been used by the majority of plants for various applications, the B&WOG's J, model was based on data from B&W-fabricated vessels only. The staff performed an extensive study, as documented in Reference B.3, and concluded that when the 1/4T fluence exceeds 10n/cm 2(E>1 MeV), the B&WOG's J, modelis less conservative than the Eason model. After examining the 1/4T fluence data for all Linde welds from the five RPVs in the current topical report, the staff found that the fluence for the most limiting weld, SA-1526, at 48 EFPY is 0.655 x 10n/cm .r At this fluence value, the staff has determined, based on the conclusions documented in Reference B.3, the B&WOG's J, model is either equivalent or more conservative than the Eason model. Hence, using the B&WOG's J, model in this application is appropriate.

Conclusion As described above, the staff has reviewed topical report BAW-2275 (Reference B.1), and determined that the B&WOG's analytical results satisfy the acceptance criteria of Appendix K of Section XI of the ASME Code. Hence, the Linde 80 welds of the five RPVs have margins equivalent to those of Appendix G of Section XI. The staff has also examined the recent best-estimate chemistry data from Framatome and concludes that the recent data have a negligible impact on the results and conclusions made in this topical report. Further, the staff concludes that using the B&WOG's J, model in this application is equivalent to using the Eason model because these two models converge to the same J, value around the fluence value of 0.655x10

2 n/cm projected at 48 EFPY for the limiting weld.

References B.1 BAW-2275," Low Upper-Shelf Toughness Fracture Mechanics Analysis of B&W Designed Reactor Vessels for 48 EFPY," Babcock & Wilcox Owners Group, August 1996.

B.2 Letter dated March 29,1994, from Brian Sheron, NRC, to B&WOG Chairman, George Lehmann, " Acceptance for Reference of Topical Report BAW-2192P, Revision 1: Low Upper Shelf Toughness Fracture Analysis of Reactor Vessels of B&W Owners Group Reactor Vessel Working Group for Level A & B Conditions."

B.3 Letter dated March 29,1994, from Brian Sheron, NRC, to B&WOG Chairman, George Lehmann, " Acceptance for Referencing of Topical Report BAW-2178P: Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Level C & D Service Loads."

B.4 Letter dated June 30,1997, from D. L. Howell, Frematome Technologies, Inc.,, to B&WOG Reactor Vessel Working Group (cc to NRC), regarding " Chemical Composition Data for Linde 80 Welds."

36

s-APPENDIX C J

ASSESSMENT OF TOPICAL REPORT BAW-2274 j

" Fracture Mechanics Analysis of Postulated Underclad Cracks in B&W Designed Reactor

{

Vessels for the Period of Extended Operation" Introduction Topical Report BAW-2274 (Reference C.1) discusses underclad cracks _beneath austenitic stainless steel weld cladding. Underclad cracks were Crst discovered in October 1970 during examination of the Atucha reactor vessel. They have been reported to exist only in SA-508 Class 2 reactor vessel forgings manufactured to a coarse grain practice and clad Ly high-heat-input submerged are processes. The underciad cracks were detected from cutouts such as nozzle cutouts from a reactor pressure vesse! forging. The regulatory position regarding this issue can be found in Regulatory Guide (RG) 1.43, " Control of Stainless Steel Weld Cladding of l Low-Alloy Steel Components. RG 1.43 states that detection of underclad cracks *normally l requires destructively removing the cladding to the weld fusion line and examining the exposed base metal either by metallographic techniques or with liquid penetrant or magnetic particle testing methods." The maximum crack size reported by the industry was 0.5 inch in length and 0.165 inch in depth. RG 1.43 concluded that the subsurface location and size of the underclad l cracks made them relatively insensitive to detection using nondestructive examination methods.

RG 1.43 did not discuss whether any of the underclad cracks were found by nondestructive j examination methods. '

B&W examined eight nozzle cutouts from forgings from B&W fabricated reactor pressure vessels (RPVs) and found the dimensions of underclad cracks in these cutouts to be less than 0.1 inch in depth and 0.5 inch in length. The details of this investigation were documented in topical report BAW-10013, dated December 1971. The RPVs that were affected are Arkansas Nuclear One, Unit 1 (ANO-1), Oconee, Unit 1,2, and 3 (Oconee-1, -2, and -3), and Three Mile Island, Unit 1 (TMi-1). In BAW-10013, the B&WOG also presented a fracture mechanics analysis to justify the continued operation of these units for 32 EFPY with the underclad cracks in the RPVs. The staff accepted the topical report in 1972. Since 1972, fracture mechanics analysis has been improved significantly. To reflect this improvement, the B&WOG employed the latest fracture toughness information, applied stress intensity factor solutions, fatigue crack growth correlations for SA-508 Class 2 material, and the IWB-3612 acceptance criteria in Section XI of the ASME Code (the ASME criteria) to evaluate the acceptability of the underclad cracks for the period of extended operation at 48 EFPY. It should be emphasized that underclad cracks are detected flaws, not postulated flaws as in the case of low USE, and therefore are subject to the ASME acceptance criteria for flaw evaluation.

The B&WOG's Assessment of Underciad Cracks The B&WOG's fracture mechanics analysis started with an assumed initial flaw depth of 0.353 inch, which corresponds to the sum of the maximum detected flaw depth of underclad cracks (0.165 inch) and the clad thickness (0.188 inch). It should be noticed that the B&WOG conservatively used the maximum detected crack cepth reported by the industry. The loading included the specified transient and external loading such as the operating-basis earthquake (OBE), the safe-shutdown earthquake (SSE), the loss of-coolant accident (LOCA), thermal 37

expansion, anc Meadweight. Cladding effect was also considered. The B&WOG used the stress profiles due to pressure, thermal loading, geometric discontinuities, and different thermal properties of cladding and nonle materials to calculate the individual appMd stress intensity factors (applied K's) for an intemal semi-elliptical surface flaw under each loading. The B&WOG then obtained the applied K by summing these individual applied K's. A fatigue crack growth-analysis under design transients was then performed to predict the final crack depth at 48 EFPY (with transient cycles corresponding to the period of extended operation). In the fatigue analysis, the applied K was first calculated. The stresses used in the applied K calculation were obtained from two finite element method (FEM) models: a reactor vessel FEM (RV FEM) model for calcula+ing stresses due to discontinuities such as those on the closure head-to-head flange and a uniform-thickness cylindrical beltline FEM (BL FEM) model for calculating stresses for vessel walls remote from any discontinuities. The Raju-Newman solutions for cylindrical vessels (Reference C.2) were used for axial flaws in the beltline and nonle belt regions, aiid liie Raju-Newman solutions for flat plates (Reference C.3) were used for axial flaws in the closure head region. The stress intensity factor solutions by Kumar (Reference C.4) were used for circumferential surface flaws.

To complete the fatigue analysis, the B&WOG calculated the difference between the maximum and the minimum applied K's, and used the ASME fatigue crack growth law to calculate the crack growth. This process was repeated and the crack length was revised until all cycles and transients had been exhausted. The B&WOG reported the maximum crack growth to be 0.18 inch, which occurs in the nonle belt region and corresponds to a final crack length of 0.533 (0.353 + 0.18) inch. The applied K due to this final crack depth was then computed for Level A, B, C, and D loading conditions.

To apply the ASME acceptance criteria, the irradiated fracture toughness is needed, which requires the estimation of the reference temperature (RTuot) of the material at the deepest point of the crack (the crack tip). The B&WOG used fluence values for the beltline region at 48 EFPY to estimate the fracture toughness properties for three regions of each RPV: beltline, nonle belt, and closure head. The crack-tip RTum values for materials in these regions were calculated using RG 1.99, Revision 2, and are documented in Calculation Summary Sheet 32-1240132-03.

The material with the maximum RTum value was identified as the controlling material in each region. The B&WOG then determined the crack-arrest fracture toughness and crack-initiation fracture toughness, K,, and K,c, respectively, for the controlling material from the RTum value using the ASME Section XI fracture toughness curves. An upper limit of 200 ksi(in)S was set for K,i and K,c, although higher values have been reported.

The last step involves applying the ASME acceptance criteria. The report concluded that when the applied K's for the final crack length were compared with K,, and K,c for different loading conditions using the ASME criteria, a margin of 3.63 was obtained for Level A and B loading and 2.42 for Level C and D loading, exceeding the margins specified in the code (3.16 and 1.41, respectively).

The Staffs Assessment on Underclad Cracks The B&WOG's methodology in performing the flaw evaluation is adequate because it is consistent with the well-established flaw evaluation procedure in the ASME Code. However, some of the technical details in the methodology have not been defined in the code, leaving room 38

for the B&WOG to develop its own approaches in these areas. As described below, the staff examined these approaches closely in this review.  ;

1 The Use of the FEM Results i i

The B&WOG used a complete vessel FEM model (RV FEM) and a beltline FEM model (BL FEM) to generate the stresses for applied K calculations. The staff has examined the details of using the results from RV FEM and BL FEM models in the analysis and determined that these models were correctly implemented. Specifically, the staff confirmed that it is acceptable to use the full vessel model (RV FEM model) subjected to normal-operation heatup, steady state, and normal-operation cooldown with rapid depressurization to identify the highly stressed regions in the vessel and to derive the discontinuity stresses for other transients. The staff also confirmed that there is negligible loss of accuracy using the partial vessel model (BL FEM model) for the beltline region because this portion of the vessel wall is remote from discontinuities. Therefore, the staff determined that the B&WOG's approach of using two FEM models in its stress analysis is 1 adequate.

Discontinuity Stresses The B&WOG provided very little information regarding discontinuity stresses in BAW-2274. In response to a staff request for additional information (RAI), the B&WOG provided extensiv e information in its proposed BAW-2274 revision for staff evaluation. The worst discontinuit) stresses (i.e., the difference between the stresses near the geometric discontinuity and the stresses remote from the discontinuity) were determined using the RV FEM model under the worst transient (pressure and thermal loading). To assess the discontinuity stresses due to l pressure alone, the B&WOG developed a reference pressure load case. It then applied a ratio of l the pressures from the worst transient to the pressures from the reference pressure load case to  !

evaluate the discontinuity stresses due to the pressure loading of the worst transient. The l discontinuity stresses due to the thermal loading of the worst transient were then determined by  ;

subtracting the discontinuity stresses due to the pressure loading of the worst transient from those due to the worst trtnsient (pressure and thermal). For all other loading conditions considered in the fatigue analysis, the discontinuity stresses were estimated by applying appropriate ratios to the discontinuity stresses due to pressure and the discontinuity stresses due to thermal loading, and toen adding them. The staff concludes that this is an acceptable engineering approach consistent with the principle of superposition in elasticity and will not compromise the accuracy of the results.

Claddino Stresses The B&WOG's methodology in calculating cladding stresses was also provided in the proposed BAW-2274 revision. The Timoshenko equation for a one-dimensional, single homogeneous

material was used to estima'e the stress across the vessel wall. The B&WOG calculated the ratio of Young's modulus for the base metal to Young's modulus for the cladding and the ratio of the coefficient of therrral expansion for the base metal to the coefficient of thermal expansion for

! the cladding. The cladding stresses were then estimated by applying these ratios to the stresses I

at the cladding location from the Timoshenko solution. The B&WOG then calculs.ted the K.

value from the differerce of the applied K values for the stress profiles with and without cladding.

The staff considered the use of a one-dimensional model a simple approximation. To 39 i

L

m ]

., o demonstrate that the simple approximation is conservative, the B&WOG compared results from this approach with those from an exact approach (Reference C.5) by Oak Ridge National Laboratory (ORNL) for a typical Normal and Upset transient and found the B&WOG's approach is 11% conservative. Based on this conservatism and the axisymmetry of the RPV and the transients, the staff accepts the B&WOG's methodology in predicting cladding stresses.

The Consideration of a Non-Desion Basis Transient As mentioned, the topical report considered design basis transients only To demonstrate that pressurized thermal shock (PTS)is not a concem for underclad cracks, the staff requested the B&WOG to perform a deterministic analysis using the extended high-pressure injection (HPI) transient from the PTS study of 1982 (Reference C.6). The B&WOG performed the analysis and reported that the ratio of Kuto applied K is 1.3 when the same set of assumptions as in the PTS study was adopted. The staff is satisfied with this ratio (margin) and concludes that neither the

)

design basis transients nor the non-design basis transients will challenge the integrity of the  !

vessel. The limiting RTers values for ANO-1, Oconee-1, Oconee-2, Oconee-3, and TMI-1 forgings at 48 EFPY are 90'F,136*F,113'F,175'F, and 127'F, respectively. All are below the screening criteria specified in 10 CFR 50.61.

Crack Growth Parameters The crack growth mechanism considered by B&WOG was fatigue. The fatigue growth parameters were from Figure A-4300-1 of Section XI of the 1989 Edition of the ASME Code. The growth rate selected was that for surface flaws in a water reactor environment. Although other

! growth mechanisms such as stress corrosion cracking (SCC) might exist, in view of operational cxperience with PWR vessels, the staff determined that fatigue is the dominate mechanism.

t Other Conservative Assumotions Besides the conservatism inherent in the flaw evaluation process, the B&WOG has introduced i additional conservatism. By using applied K equations for surface defects in the underclad crack fracture mechanics analysis, the B&WOG has built in additional conservatism in the evaluation process. Also, a conservative number of cycles,360 (6 heatups and 6 cooldowns per year), was l l cssumed for the 100'F /hr normal heatup and cooldown transient. l l

Conclusion As stated previously, the B&WOG's methodology in performing the flaw evaluation is consistent with the current well-established flaw evaluation procedure and criteria in the ASME Code, and, therefore, is adequate. The staff also examined the unique aspects of the, B&WOG approach cnd found them adequate. The additional conservatism associated with the B&WOG's methodology includes: (1) using the maximum crack depth of 0.165 inch reported by the industry cs the initial crack depth instead of the depth of 0.10 inch reported on the B&W RPVs, (2)

Essuming all underclad cracks are surface cracks, (3) using the fatigue crack growth rate for surface flaws in a water reactor environment, and (4) producing results equivalent to a safety factor 17% more than that specified by the ASME Code for Level A and B loading and 72% more than that for Level C and D loading.

40

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References C.1 BAW-2274, " Fracture Mechanics Analysis of Postulated Underclad Cracks in B&W Designed Reactor Vessels for the Period of Extended Operation," Babcock & Wilcox Owners Group, December 1996.

C.2 1. S. Raju and J. C. Newman Jr., " Stress-Intensity Factors for Intemal and External Surface Cracks in Cylindrical Vessels, " Joumal of Pressure Vessel Technology, ASME, Vol. 104, November 1982.

C.3 J. C..Newman Jr. and I. S. Raju, "An Empirical Stress Intensity Factor Equation for Surface Cracks," Engineering Fracture Mechanics, Vol. 15,1981.

C.4 V. Kumar, M. D. German, and B.1. Schumacher, " Analysis of Elastic Surface Cracks in Cylinders Using the Line-Spring Model and Shell Finite Element Method," Journal of Pressure Vessel Technology, ASME, Vol.107, November 1985.

C.5 J.A. Keened and T.L. Dicks on, " Stress-Intensity-Factor influence Coefficients for Axially Oriented Semi-elliptical Inner-Surface Flaws in Clad Pressure Vessels (R/t=10)," Letter Report ORNL/NRC/LT.-93/33, Revision 1, prepared for U.S Nuclear Regulatory Commission, Oak Ridge National Laboratory, Oak Rid 9e, Tennessee, September 30,1995.

C.6 Policy issue from J. W. Dirks to NRC Commissioners,

• Enclosure A: NRC Staff Evaluation of Pressurized Thermal Shock, November 1982." SECY-82-465, November 23,1982, Division of Nuclear Reactor Regulation,'U. S. Nuclear Regulatory Commission, Washington, D. C.

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